Text

Advises of Completion of Reactor Sys Branch Review of Info Presented in P Crane & 771122 Meeting.Applicant Has Adequate Sys for Bringing Plant to Cold Shutdown in Event of SSE
	ML20214R337
Person / Time
Site:	Diablo Canyon
Issue date:	03/08/1978
From:	Ross D; Office of Nuclear Reactor Regulation
To:	Vassallo D; Office of Nuclear Reactor Regulation
Shared Package
ML20213D572	List: ML20213D570; ML20213E944; ML20213F093; ML20213F114; ML20214R269; ML20214R304; ML20214R337; ML20214R407; ML20214R489;
References
	FOIA-84-741, FOIA-84-742 NUDOCS 8706080086
	Download: ML20214R337 (27)
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"I NUCLEAR REGULATORY COMMISslON A

WASHINGTON. D. C. 20065 s-/

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l Docket Nos. 50-275/323 MAR U 1978 I

4 MEMORANDUM FOR:

D. B. Vassallo, Assistant Director for LWR's, DPM l

FROM:

D. F. Ross, Jr., Assistant Director for Reactor Safety, DSS

SUBJECT:

COLD SHUTDOWN HEAT REMOVAL Plant Name:

Diablo Canyon. Un H Nm 1 and 2 l

Docket Nos.:

du-us/ 323

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Licensing Stage:

_m Requesting Branch LWR-1 1

and Project Manager:

D. Allison Systems Safety Branch Involved:

Reactor Systems Requested Completion Date:

March 1, 1978 j

Review Status:

Complete i

The Reactor Systems Branch has reviewed the information presented in i

letter from P. Crane (PGaE) dated January 26, 1978, and at a meeting with-3 l

the applicant on November 22, 1977.

Based on this review, we find that the applicant has adequate systems for bringing the plant to cold shut-(9 down in the e. vent of an*SSE as-discussed in the enclosure.

It should i

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be noted that this conclusion is based on the assumption that the

Auxiiiary Systems, Instrument and Control Systems, and Mechanical j

Engineering Branches will confirm the adequacy of those aspects of i

achiaving cold shutdown that are under their cognizance.

In particular, the hookup of the auxiliary feedwater system to the raw water reservoir and auxiliary saltwater system, the component cooling water system, and operation of the steam generator relief valves must be evaluated by ASB.

ICSB needs to review operation of all components noted in our i

i memo of December 8,1977 and also has to.confinn that position indication is provided on all essential hydraulic or electric operated valves.

Finally, the MEB must confirm that all of the equipment meets (SE 1

l requirements, i

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D. 'F. Roos, Jr. ' Assistant Director.

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DEVINE84-741 PDR Divi.sion of Systems Safety cc:

5. Hanauer T. Ippolito R. Mattson-R. Tedesco c

i D. Ross J. Knight l

J. Stolz.

T. tfovak

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V. Benaroya S. Israel i

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Contact:

S. Israel, NRR

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(q 5.4 Cold Shutdown Heat Removal The applicant has evaluated his plant with respect to achieving long term cold shutdown conditions in the event of a Safe Shutdown Earthquake (SSE) which also results in an extended loss of offsite power.

For this event,. the basic functions that must be performed following scram arer a) heat removal b7 steant generator depressurization c) primary system boration i

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primary system depressurization e

surveillance of boron concentration f

surveillance of primary and secondary system parameters The applicant has stated that all, systems and equipment needed to t

perform these functions will be qualified for the SSE and the staff's, review of the seismic qualification of the equipment is presented in Section Initial heat removal wili be through the steam generators using the auxiliary feedwater system and releasing steam to the atmosphere by way of the safety valv4s The condensate storage tank is the primary source of feedwaten.with about an 8 hour9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months water. supply, In order ta ensure capabil-ity to remove ~ heat via the steam generators-for extended periods, proirisions have been made to connect the raw water reservoir (minimum T,000,000 gallon unit) to the suction line of the auxiliary feedwater pump.

This will provide an addi-tional 100 hours0.00116 days 0.0278 hours 1.653439e-4 weeks 3.805e-5 months of steam generator operation'for both units. As a further backup, design provisions allow sea water from the 4

auxiliary salt water system to be used as auxiliary feedwater makeup l

4 if necessary.

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'After the primary system has been cooled down to 350'F and de-pressurized to 425 psi, the Residual Heat Remval (RHR) system is activated and the decay heat is transferred to the ultimate heat l

sink via the component cooling water (CCW) and auxiliary saltwater systems.

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The steam generat' ors are depressurized to cool down the primary.

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system. Air-operated relief valves are used to depressurize the steam generators. The primary system is depressurized by spraying

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l cold borated water into the pressurizer.

This is accomplished i

with an auxiliary spray line connected to the charging pumps in.

the CVCS.

j Boration of the primary system is performed with the CYCS,using i

1 two boric acid tanks having sufficient 12 W/o boric acid to borate i-l the reactor coolant system for cold shutdown conditio~ns.

Provisions are available for taking periodic ' samples of. the reactor coolant l

g tar conff m the baron concentration.

j Normal instrumentation is avaiTable to maint'ain surveillance of i

important primary and secondary system parameters such as' pressure,,

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temperature, and water levels.

i The applicant has presented an anplysis 'of the primary system water.

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volume shrinkage due to cooldown and has determined that the shrinkage is sufficient to accommodate CVCS input required for boration and cooling of the pressurizer.

Therefore, the normal j

letdown system is not required to achieve long term cooling with i

RHR system. Similarly, the applicant has demonstrated that de-gasification of the reactor coolant is not necessary because'the t

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potential H2 ccncentratioq in the solution is less than the i

3 saturation value at RHR operating conditions.

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ATT of the operato'r actions needed to perform plant cooldown (except for periodic surveillance:of the boron concentration) can be accomplished from the control room assuming no single failure.

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The applicant has demonstrated that redundant' paths or systems i

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. are available to perform the essential functions in the event of j

a single failure, however, in 'some instances this would require operator action outside of the control room to activate the redundant j

path.

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The RHR system har 'a single suction line with two isolation valves

'L in series. The staff was concerned about a possible mechanical failure of one of these valver which would preclude activation

,of the RHR. system. The applicant har indicated that the mean mechanical failure rate for this type of valve is 5 x 10-4 per year.

This mechanical failure rate combihed with.the probability of having -

1 an earthquake largar than an 08E (7 x 10-3) is about 10% of the risk of core melt from all causes calculated in the Reactor Safety j

Study (WASH-1400)

It should be noted that long ters heat removal,

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- feedwater system if the RHR system cannot be activated. As a result, the staff considers the probability of this failure and its potential consequences to be acceptable.

The applicant is reviewing all of non-Class I lines inside containment to confirm that the postulated failure of any such line would not interfere with safe shutdown.' The staff will report its evaluation of the applicant's rev,iew in a supplement.

The applicant has committed to performing a natural circulation test to demonstrate adequate baron mixing and acceptable cooldown conditions under natural circulation. These tests would be performed during startup when core heat is available for these tests.,

The staff has reviewed the applicant's capability to cool down his i

plant to RHR conditions and provide long. term cooling on the RHR.

t He has demonstrated that sufficient systems are available for residual heat removal with or without offsite power and assuming i

a single failure as required by General Design Criterion 34 and is therefore acceptable.

Similarly, these systems can be operated in the event of a SSE as required by GDC 2 and discussed in 1

Section pending review of potential adverse failures of j

non-CTass L equipment insjde containment.

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7. 0 INSTRUMENTATION.AND CCNTROL

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7. 2 Reactor Trio System In Supplement Nunber 7 to the Safety Evaluation Report we found the basic seismic scram system proposed by the appiteant acceptable.

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However, we required further information from the appifcant regarding how the system would satisfy our requirements for separation, isolation quality, testability.,

and qualification for Class 1E circuits.

The applicant has provided additional Information on this sutject in Amendment 67 i

to the Final Safety Analysis Report. Based on our review of the additional h

information, we have concluded that the seismic scram system is of a similar design 1

and meets the same criteria as the reactor protection system and is, therefore, d

acceptable.

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We consider this matter resolved.

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7. 4 Systems Required for Safe Shutdown u.>

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In Supplement Number 7 we stated that we would require further information about

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the indication available to the control room operator in connection with performing i

a shutdown after a Hosgri event. We have now completed our review of this matter and it is resolved as discussed in Section 3.2.1 of this supplement.

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7. 6 RHR Overpressure Protection Interlocks

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In the Safety Evaluation Report we described the interlocks for the motor operated y

valves on the residual heat removal system suction lines (Valves 8701 and 8702).

The interlocks operate on diverse principles to prevent opening the valves when j

1 reactor coolant system pressure exceeds about 425 pounds per square inch and to automatically close the valves when reactor coolant system pressure exceeds about 600 pounds per square inch. The interlocks are provided to prevent overpressuriza-tion of the residual heat removal system when reactor coolant systes pressure is j

high, primarily during operation. We found these interlocks acceptable.

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In our fire protection review it was postulated that fire damage to electrical 3

cables could cause both valves to open. To correct this the applicant has propes e removingpowerfromthival"ves'motorcperatorsbyopeningmanualcircuitbreakers.

Since this wtf1 prevent the postulated fire damage from opening both valves we i

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consider it acceptable. However, to provide continuous indication that the valves are in the correct position, we will also require installation of redundant control room valve position indication that is not disabled by opening the circuit breakers.

l Alternately, the applicant may orovide further separation of the electrical cables.

J Separation that meets our fire protection criteria would also be an acceptable y

method of resolving the fire protection question. It would eliminate the need to remove power from the valves and to install redundant valve position Indication.

e The applicant has agreed to provide one or the other of the acceptable modifications y.

described above. We will require appropriate documentation, 6,

Y Based on the applicant's comitment to meet our requirements we consider this

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matter resolved.

We will confirm compliance with our requirement for documentation prior to issuance of operating Ifcenses.

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1 7.6 RHR isolation Valve Position Indication 4

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.In SER Supplement 8 we postulated that fire damage to electrical cables could cause both RHR

) _ y suction Ifne isolation valves to open. The applicant has since comettted to remove power.

l 59 from the valve motor operators to prevent this unoesired operation. We find this response as l g acceptable.

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7.8 Environmental and Seisele Qualification i

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Environmental Qualification of Submerced Equipment 1

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We stated in SER Supplement 7 that we had identified the RCS Pressure, Pressurizer Level, s

I5 and Marrow Range Steam Generator Level Transmitters as being vulnerable to malfunction due i

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to submergence and that the applicant had agreed to replace them with qualified instruments j

i'.T or to, justify their continued use. We have reviewed additional information that presents

'k the results of submergence testing of the replacement Sarton transmitters. This infor1eation j

C demonstrates that these transmitters will perform suisfactorily when submerged. We find y,

this response acceptable.

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Environmental Qualification - Nuclear Steam Supply System Eaulosent Exposed to Normal Environments

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i Supplement 7 to the SER stated that we would complete our evaluation of temperature ranges and of the ambient temperature monitoring system in a future supplement..The ambient tempera-ture monitoring system that monitors the areas containing 80P safety-related equipment will i

also monitor areas outside of containment that house MSS $ safety-related equipment which exceed, with adequate margin, the design basis controlled temperature range of the environ-3 mental control systems. Should the ambient temperature of given areas exceed this design.

l basis controlled temperature range, there would be ample time to evaluate the affected safety' equipment's capability for continued service and take corrective action, if required.

We find this acceptable.

Environmental 0ualification - Class IE Equipment Ex' posed to Severe Environments

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In SER supplement 7 we required the applicant to furnish a listing of all 80P and NSSS safety-related equipment that may be required to function under severe environmental condi-tions. This list was to identify the equipment, its manufacturer..its model number, its 2

location, and a specific reference to its qualification report. We reviewed PSAR-Table 3.11-1A which lists this equipment and provides the requested information.~We find this response acceptabie.

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'j NUCLEAR REGULATORY COMMISSION o,

UNITED STATES v

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. WASHINGTON. o. C. 20555

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h((N Y Sw~sc m2 MEMORANDUM FOR:

Frank '

"' ' ':, Assi stant Director for M 'et; ':h:=:d, DST f1L.

Teckfoy FROM:

R. Wayne Houston,. Assistant, Director for Reactor Safety, DSI

SUBJECT:

RCS/RHRS RCPB ISOLATION VALVES FOR PWRs By memorandum to Roger Mattson dated April 17, 1984, Tom Dunning, ICSB, raised a concern regarding the loss of an instrument bus that would cause auto-closure of the isolation valves on RHR suction droplines, with accompanying recommenda-tions for the issuance of a Generic Letter to all licensees on./the

SUBJECT:

REQUIRED ACTIONS FOR RHR SUCTION VALVE INTERLOCKS.

We have briefly reviewed the totality of the Operating Safety and krotection requirements for this valve system, see attachment 1, and recommend that a substantive review is necessary, as a generic issue, before being able to justify additional recommendations and requirements.

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We propose that the related circumstances place this generic issue primarily in Category 4d of Enclosure I of NRC Office Letter No. 40 - Management of Proposed Generic Issues:

namely:

Generic Issues that relate to NEPA [and which]

will be sent to the Division of Engineering (DE) to be: assessed and assigned a priority ranking based on their significance and cost to implement.

R. Wayne Houston, Assistant Director for Reactor Safety, Division of Systems Integration

Attachment:

As stated cc:

R. J. Mattson R. Vollmer B. Sheron J. P. Knight V. Noonan L. G. Hulman K. Kniel F. Rosa N. Lauben L. Marsh T. Dunning R. Licciardo 1

W. Minners C. Seyfrit A.T(ewberry adani H. Vandermolen S. N J. Page--

A. R. M'a chese H. Ornstein

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3Gavsonne MEMORANDUM FOR:

Frank f. "'r:f ', Assistant Director for S:f:t3 ?.;;;;am..., DST

'/Wennetegg FROM:

R. Wayne Houston, Assistant Director for Reactor Safety, DSI d

SUBJECT:

RCS/RHRS RCPB ISOLATION VALVES FOR PWRs By memorandum to Roger Mattson dated April 17, 1984, Tom Dunning, ICSB, raised '

a concern regarding the loss of an instrument bus that would cause auto-closure of the isolation valves on RHR suction droplines, with accompanying recommenda-tions for the issuance of a Generic Letter to all licensees on the

SUBJECT:

REQUIRED ACTIONS FOR RHR SUCTION VALVE INTERLOCKS.

We have briefly reviewed the totality of the _ Operating Safety and Protection requirements for this valve system, see attachment 1, and recommend that a substantive review is necessary, as a generic issue, before being able to justify additional recommendations and requirements.

We propose that the related circumstances place this generic issue primarily in Category 4d of Enclosure I of NRC Office Letter No. 40 - Management of 1

Proposed Generic Issues:

namely:

Generic Issues that relate to NEPA [and which) will be sent to the Division of Engineering (DE) to be assessed and assigned a priority rankin-based on their significance and cost to implement.

R. Wayne Houston, Assistant Director for Reactor Safety, Division of Systems Integration

Attachment:

As stated cc:

R. J. Mattson R. Vollmer B. Sheron J. P. Knight V. Noonan L. G. Hulman K. Kniel F. Rosa N. Lauben L. Marsh T. Dunning R. Licciardo

. W. Minners C. Seyfrit A. Thadani H. Vandermolen S. Newberry J. Page - -

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R. Marchese H. Ornstein

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.D:cument Name:

GENERIC ISSUES Recuestor's ID:

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Author's Name:

Licciardo A Document Comments:

6/8/84 - DRAFT / Please return this sheet with revisions 5

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Attachment No. 1 PROPOSED GENERIC ISSUE RCS/RHRS RCPB ISOLATION VALVES FOR PWRs 1.

Suggested Title of Proposed Generic Issue or new requirement.

RCS/RHRS RCPB Isolation Valves [ Reactor Coolant System / Residual Heat Removal Suction, Reactor Coolant Pressure Ecundary Isolation Valves] For PWRs [ Pressurized Water Reactors]

2.

What is the known, suspected, or potential deficiency in the technical basis of existing staff guides or requirements?

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Suso~ect of a deficiency A deficiency is suspected because of the difficulty being experienced in establishing a generally acceptable basis for the instrumentation and control (both automatic and administrative), for these RCPB valves,

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when used under normal steady state operating conditions as an isolation. valve for the RCPB in MODES 1, 2 and 3.

And, when the valves are opened, so that the RCPB is not isolated, in MODES 4, 5 and_6, to allow for operation of the Residual Heat Re= oval System.

The difficulty lies in the inability to a-chieve acceptable levels of safety, as perceived by different entities, against the consequences of single failure in the valve system during either of the above normal operating conditions.

c.

Potential deficiency '- -

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J An incompletg. clarification of all the significant operating phases g

in MODES 1 through_6, but especially in MODES 3 through 6.

MODES 3

-- a-06/15/84 1

GENERIC ISSUES-

through 6 are now m' ore' completely and accurateiy subdivided for the purposes of operation, and necessary evaluation and protection, into

'3 phases for MODE 3, 2 phases for MODE 4, 2 phas'es for MODE 5 and 2 phases for MODE 6.

An incomplete clarification, and definition of the safety role of these valves in every MODE of normgl operation from MODE 1 through 6.

An incomplete clarification, and definition, of the protection func-tions of these valves for current Design Basis Occurrences II through IV, and Fire Protection, in MODES 1 and 2.

An inacequate clarification, anc cefinition of the pNtection func-tions of these valves for appropriate potential -Design. Basis Occur-rences (of apparently. reduced severity), and Fire Protection, in MODES 3 through 6.

An incomplete Licensing Bases may exist for the identification evaluation and acceptable protection from new transients _and ac-cidents which ar,e introduced in MODES 3 throush 6 and which are not bound by the Design Basis Occurrences, and related protections, e.g.,

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new events leading to the need for overpre'ssure protection in both non water solid and water solid circumstances are being identi#'ed; incomplete consiceration may be given to the failure of RHR in relatively critical MODES 5 and 6 with low water level (i.e., water level below the ' reactor vessel flange); T.S. which are not in con-formity with the Licensing Basis leading to unexpected loss of heat removal capability which has not been evaluated.

Reviews have raised concern for the continuing integrity of the RHR pump on inadvertent closure of the subject valves.

This could-derive primarily from an inadeouste evaluation 6f the thermai hydraulics of the instaliec syste, en single valve failure, for tne differing normal opera'tini and protection circumstancer f1rr _ttie sys-

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Q tems using the RCS/RHRS isolation valves and the RHR pumps.

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06/15/S4 2

GENERIC ISSUES t

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The protection functions of the.abovi valves are equired La.

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.f ;%: ' n ine ic ; t: - to maintain "Long-Term Cooling" in accorda :e with 10 CFR 50.46.b(5) during recirculation after'a LOCA event, c.

Known deficiency

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a)

RCPB orotection

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These two valves are Reactor Coolarit Pressure ~ Bou[dary (RCPB) valves, in'accordance with 10 CFR 50.2(v)(2)(1).

These RCPB valves are part of the Reactor Coolant Prer. ire Boundary penetrating the containment and-as~suen '.e..reo md

..o be provided with automatic isolat. ion en each valve', Ensice and outside containment, with automatic isolation being designed to take the position that provides greater safety.

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Because these valves provide. primary protection for the integrity of the RCPB and hence of the primary fission product boundary of the fuel elements, they are required to be designed, fabricated, erected, constructed, tested and inspected to quality standards

- commensurate with this importance.

Reference 10 CFR 50.55.a(a)(2)

Codes and ' Standards" specifies the necessary requirements for these valves which in general are for Class 1 of the appropriate Draft or Final ASME Code for Pumps and Valves for Nuclear Power under 10 CFR 50.55.a(f) with related inservice inspection and testing requirements to 10 CFR 50.55.a(g).

10 CFR 50.55.a(a)(2) also requires that the protection systems of nuclear power reac-tors of all types shall meet the requirements described in 10 CFR 50.a(h) which in general requires conformance to IEEE-279.'

Further, for tnis ::urpose.

o ensure maximum RCPS Integrity, the valves should be ca:able of automatic isolation to the requirements of Foottmte 2 to 10 CFR 50.55a, and 10 fFR.50 App. A,

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Criterion 55(4) [on Containment Isolation A signal], initiated by the Sa,f.ety Injection Signal.

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GENERIC ISSOES-06/15/84

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s At this time, it is the writer's observation that some of the x,,,

above requirements may not be being met, and particularly with respect to protection requirements includin'g Isolation on SI.

b)

RHR - Residual Heat Removal Use of these open valves in tbs RHR should not impinge their primary protection requirement for the protection of the RCPB.

RHR is a normal operation activity required by 10 CFR 50 to be-capable of achieving its function after allowing for a single failure.

e Principal problems arise, because it may be that contrary to Regulatory Requirements, plants have been licensed i8 which the RHR system cannot continue to perform -its function on a single active failure causing closure of one, or two, of the RCS/RHRS

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RCPB valves.

Further, in these situations, use-of alternate systems and MODES, while they may be capable of performing the necessary function, do not appear to be within the intent of the Regulatory Requirements and therefore not in acco-dance with 10 CFR 50.

Further, any attempt to rectify this situation

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by interfering with the SI protection function, must be'consid-ered as renderin5 inoperable, and a bypass of, a Safety Related Function.

c)

RHR - Overoressure protection of the RCS i)

Use of Re-Set PORVs Reduced temperatures within the RCS require' overpressure protection at pressure levels which are much less than those or0v'ded by the related 00de Safety Vaive.

For the above purpose, the NRC allows the use of PORVs from the--pressurizer which are re-set to the required low pres-sures fot Appendix G protection at RHR temperatures..One a-06/15/54 4

GENERICISSdES'

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e must questici) the validity of using eqvfpment other than

't Code Safety Valves for Overpressure Protection.

Further,

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especially when such valves may not have been qualified to RCPB Integrity Protection System requirements.

This method of overpressure protection for the RCS does not recuire the RCS/RHRS RCP3 isolation valves to be open to ensure RCS integrity.

It does, however, require limited operability for the relatively high pressure pumps of the ECCS which currently does not appear to be a part of W STS requirements which may therefore be non-conservative.

Further, a recent submittal by Westinghouse has shown that an overpressurization event - albeit capabl,e of being handled where re-set PORVs are employed, and with limited ECCS operability - is generated by the system control / pressure response of the RCS system, on closure (inadvertently or i

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otherwise) of the RCS/RHRS RCPB Isolation Valve. This is an example of the need for discovery of all potential over-pressurization events under these qircumstances of MODES 3, 4, 5 and 6.

ii) Use of Relief Valves in the RHR System The NRC has allowed the use of Code relief valves in the RHR s9 stem as overpressure protection for the RCS.

This requires that both RHR trains be available in the event that one is isolated cy inadvertent isolation of the RHR/fCS Isolation Valves.

One must again question the validity of using these relief valves not qualified to RCPB Integrity and Protection System Requirements for protection of the Nuclear Class 1 c,

RCPB of the RCS.- This is potentially contrar-y to Regulatory Requirements.

t 06/15/84 5

GENERICISSbES

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dq-Under these ci'rcumstances,. there is a mar ed conflict in Pro-N _. <'

tection Requirements.

Open RCS/RHRS RCPB IVs are essential for overpressure protection of the RCS, but any' failure of the integrity of the RHR relief valves or the RHR system leaves the RCS absent its primary protection of automatic isolation unless there is an automatic override by the SI' signal, of any' logic keeping the valves open..Addttionally, if the RCS/RHRS valves were able to close in sufficient time (for say a small break).

then there is no obvious overpressure protection for the RCS~

system immediately subsequent to the event.

Effectively:

Overpressure protection for_the RCS system using' facilities which are not of the RCPB and related protection requirements is contrary to the intent of the _ Regulatory Require-ments and cannot - as a matter of fact - ultimately p'rotect the RCPS because of the need for a continuous pressure boundary which must in fact be ultimately isolated to " protect" the RCPS if. a break occurs in the non RCPB. portion.'

s.j d)

Standard Technical Soecifications On a detailed review of Technical Specifications developed.on the basis of the W STS, the writer has found gross deficiencies in requirements for Operability when compared with the Licensing Bases, for the RHR, Component Cooling Water (CCW) and Nuclear Service Water (NSW) systems, without the required safety evalu-ations.

The writer suspects that many events of loss of RHR -

capability are associated with these gross -faults.

e)

The writer would propose, from the above examples, that it is possible that non-conformance to Regulatory Requirements has resulted in conflicting situations which may_ never have arisen if the original requirems ts nad baen more appropriately con-

'S formed to.

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c C6/15/64 6

GENERICISSdES L

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(s What present specific sa'fety requirements (e.g., SRP, R'egulatory Guide, 3.

, _j Rule) appear to be inadequate or in doubt?

We have not undertaken a detail examination of RGs and SRPs.

The principal problems seem to be in interpretation of what the engineer-ing' elements of the system are within the framework of 10 CFR 50, e.g.,

RCPB valves, requiring Protection System grade I & C, and Power Supply; Overpressure Protection for RCPB; RHR and the regulatory requirements for single failure capability within the RHR system itself rather than with other systems, which appears to be contrary to _ regulations.

4 If a new ' requirement is proposed what is the proposed requirement?

Provide, to the extent possible a value-impact assessment.

A first act'on would be to engineer, within the existing Design Basis

)

Occurrences, to the existing Regulatory Requirements for RCPB and RHR sys-f

(~ S')

tems.

Lack of this introduces fresh problems and contradictory incompatible positions.

Reference Item 2.c. under "Known Deficiencies."

/

A second action would require work on eliminating the Potential Deficiencies identified in 2.b. above.

5.

What new information must be developed either to confirm the adequacy of the current

technical bases or to define new requirements that would restore adequate protection?

i Reference second action in Item 4. above.

6.

What actions are being taken (if any) or should be taken on operating pla'nts to correct the suggested deficiency?

By whom (organization and individuai) are these actions being taken?

f

~

The writer is aware of the '~oliowing activities by the NRC,'"wblch' tan be

~-

Subject:

"RHR Interlocks for Westinghouse Plants." Memo for R. Mattson thru R. Wayne Houston from T. Dunning, Section Leader,.

ICSB to,the above subject and dated April'17, 1984.

Enclosures I through 4.

f)

Proposed Generic Letter:

Concerning-Recuirec Acticns.for RHR Suction Valve interlocks.

.cre;osai cy Tem Dunning. dated approximately May s

r

1934,

- - ~.,

t 1: -

t 06/15/84 5

GENERIC-ISSbES

~

9

, 3 ' e) and f) The current action 'evoived from a request by Tom Dunning, ICSB,- items i

e N,/

e) and f) above, arising from his consideration of the work performed

'by AE00 described _under d above.

In this request from ICSB, it was s

proposed to issue a " generic letter" requiring licensees to review the engineering of their RHR valve interlocks in accordance with what is effectively a guideline, to be followed by a report and pro-posed necessary (as appropriate) amendments.

It is perceived that the nature of the action was more in accordance.with clarifying or correcting ' current requirements or guidance and not requiring changes to the SRP or a Regulatory Guide.

In which case the processing:and resolution we'uld become the responsibility of the appropriate division and not a part of the procedure in Enclosure 1 of NRC Office Letter No. 40 dated March-19, 1983.

g)

The issue is als'o related to the general subject of "Over pres'sure Protection of the RCS System," which is treated as an element of review in the SRP NUREG-0800, Branch Technical Position RSB 5-2,-

[

)

"Overpressurization Protection of Pressurized Water Reactors While Opersting At Low Temperatures."

It is also related to the general subaect of Overpressure Protection of the Rescidual Heat Removal System which is treated as an element of review in the SRP NUREG-0800 under Branch Technical Position RSB 5-1, " Design Requirements of the Residual Heat Removal System."

h)

The issue must be considered as related to a recent W submittal for-Topical Report Review of: WCAP 10529, COMS COLD OVERPRESSURE MITIGATING SYSTEM, Westinghouse Electric Corporation, Febuary 1984.

Non NRC Activit"ies:

A major activity has also been undertaken in this area by the Nuclear Safety Analyses Center operated by the Electric Power Research Institute.

inis is entitled:

" Residual Heat Removal Excerience O

ReviewandSafetyAnaiyse$,PressurizedWaterReactors N$AC/52, January 1983".

This document provides eloquent experience on loss of coolant a'crfdents occurring in MODE 5, as well as on Loss of RHR j

06/15/84 9

GENERICISSbES j

In respeci to loss of coolant acciden s, page 2-2 of the (O

events.

a

\\

/

document states "RHRS isolation would have halted the inventory-

'. loss in every cold shutdown loss of coolant everit in the last five years."

7.

If the i'ssue is related to another generic issue (e.g., TMI Action Plan Item), identify the generic issue and the area of issue overlap.

The issue is related to item 6(a)-of this review, namely USI A-45, of which it is an element.

This USI, entitled, " Shut Down Decay Heat Removal.

Requirements" examines resolution of Loss of Decay Heat Capabil-ity in MODE 3, as well as MODES 4 through 6, by means which include those outside the current Design Basis.

The writer proposes that the current _ status of the RCS/RHRS IV problem is one which should be capable,of generic reso-lution within the existing Regulatory Basis, and'in doing se ca' bo'th draw n

upon, and contribute to, the w;rk of A-45.

(

The issue is also related to item 6(c) of the proposed generic review, D

"RHR Suction Valve Testing." The RCS/RHRS RCPB isolation valves are also valves, which on their failure open, could cause en ' Interfacing system

/

LOCA through containment.

Given appropriate priority, this task will analyze the importance of inspection in establishing the probability of structural ftilures of these valves resulting in such circumstances, using resulting risk estimates.

For this proposed generic issue on RCS/RHRS IVs, an important element of concern will be the probability of inadvertent

" failures" in the I&C and Electrical systems finally proposed for these valves (after deterministic solutions) causing and/or sustaining such Interfacing fystem LOCA circumstances during their various operational MODES.

Failure to remain open under normal operation 'in the RHR mode, and failure to remain closed under accident conditions during the recirculation mode, including containment spray, also have related risks which would also be analyzed.

At this time it is suggested that the methodology of this pro-posed generic task on RHR suction valves testing could be directly appli-1 O

cable to assessing some of Ihe'most important risks deriving f~r'om'in-advertent failures to close, or remain closed, of these RCS/RHRS~RCPB i solation valves;.-~

06/15/84 10 GENERICISSbES l

~

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O s

The proposed issue is al'so related to another generic fssue under s

7

(

prioritization, namely, Generic Issue No 24 - Automatic Switchover to Recirculation. The issue is currently with the Generic Issues Branch for prioritization under Task Manager Harold T. Vandermolen. This issue is assigned to RSS, and the writer R. B. A. Licciardo, for resolution on establishment of Priorities. The-relationship between this and the pro-posed issue, is that re-alignnentof ECC1 for recirculation necessitates continuing integrity of the RCPB isolation valve function of the RCS/RHRS valves under consideration, as well as the important related issue of main-taining RHR pump integrity to effect continuing ECCS cooling both directly, and indirectly in its role in providing necessary NPSH for SI and C.C.

Pumps.

The RHR pumps are also necessary to continue containment spray in the recirculation mode where the RHR pump is designed to so provice (as in Salem 1 & 2), and thereby are essential for contain. ment integrity.

8.

Is anyone currently working on this issue?

If so, name and organization.

('- '

Except for the action taken by ICSB discussed in 6.e. and 6.f., there is no one working on this particular RCS/RHRS IV item at a proposed generic issue.

9.

References See earlier references under item 6. above, as a starting basis.

==

Conclusion:==

1)

The multiple duties required of those RCS/RHRS RCPB Isolation Valves in-clu'de:

RCPB protection, and especially from an Interf acing System LOCA.',

during normal operation in all MODES; RCPB Isolation on an SI signal: RCPB protection sucsecuent to-any other Licensing Basis Event in the Short Term;

~ ~

and in the Long Term for decay heat removal during re-circuTation for small

~

\\,,,/

and large break LOCAs including core and containment spray cooling; and potentially Conta3nment Spray Coeling on main steam line break in

'#~

06/15/84 11 GENERICISSbES

q containment. The duties of'the same valves also include being opened for:

k Single failure proof RHR operation in MODES 4, 5, and 6, and overpressure protehtien for the RCS via RHR relief valves for a limited numoer of plants.

2)

The principal difficulty lies in the recognition of.the multiple duties recuired'from these valves under normal operating, transient and accident conditions in the short run, and acciden_t conditions in the long run.

The different nature of each of these circumstances demands different sets of requirements from within 10 CFR. Solution to the difficulty lies in an accurate identification of those functions and conditions to which these valves are to be subject, a listing of the related Regulatory Requirements, an assurance that the engineering resolutions accepted will not conflict, and that Regulatory Requirements conforming to the most stringent safety functions required for each of the particular system elements be adopted.

S)

The importance of the Protection and Operating Safety Functions may vary between Facilities because of differences in ECCS, RHR and Containment Spray Cooling provisions.

I 4)

At this time, it is proposed there is a reasonable probability that a

/

generic approach for PWRs based on deterministic solutions can be achieved within the current Regulatory Framework, with associated potential modi-fications to SRP and RGs, arising from the work of 2) above.

5)

The approach adopted should review all the information included [and ref-erenced] in this attachment, in assessing the-potential revision of staff guidance and requirements arising out of existing regulations.

This revi-

~

sion can then be used as a basis for achieving deterministic solutions, within such regulations, which may have potent'ially a broad basis of appli-cation across th. different sets of existing engineering. And later, as priorities demand, for new Licensing Acplications.

6)

Given an number of deterministic si..tions, it is proposed that Risk 3

Assessments can then be used to evaluate the most cost ef fectiv'e..If

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06/15/84 12 GENERIC ISS ES

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.4 thisworkshowsthattheproposedsolutionsa)enotacceptableinall cases,.then proposed changes to the Regulations may then be considered.

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7)'

RSS proposes that this review plack this generic issue, primarily in Category 4d of enclosure 1 of NRC office letter No. 40 Management 0f Proposed Generic Issues:

namely " Generic Issues that relate to NEPA

[and which) will be sent to the Division of Engineering (DE) to be assessed and assigned a priority ranking" based on their significance and cost to implement.

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13 GENERICISS0ES 9

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Olan Parr, Chief Auxiliary Systems Branch Division of Systems Integration Faust Rosa, Chief Instrumentation & Control Systems Branch Division of Systems Integration

' ' ~ - 2 :...., :.'.. J Reactor Systems Branch Division of Systems Integration i

Vincent Noonan, Chief Equipment Qualification Branch Division of Engineering George Lear, Chief i

Structura.1 and Geotechnical J

Engineering Branch

^

Division of Engineering (j-7 Robert Bosnak, Chief Mechanical Engineering Branch Division of Engineering Ashok Thandar,i, Chief Reliability & Risk Assessment Branch Division of Safety Technology

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