NUREG-0422, Requests Comments on Review of Encl marked-up Tech Spec Concerns Resulting from Differing Prof Opinion.Some Concerns Should Be Resolved on plant-specific Basis.Proposed Amend Requests & Reply Needed within 90 Days

Requests Comments on Review of Encl marked-up Tech Spec Concerns Resulting from Differing Prof Opinion.Some Concerns Should Be Resolved on plant-specific Basis.Proposed Amend Requests & Reply Needed within 90 Days

ML20128N380
Person / Time
Site:	Mcguire, McGuire
Issue date:	07/09/1985
From:	Novak T Office of Nuclear Reactor Regulation
To:	Tucker H DUKE POWER CO.
Shared Package
ML20127A476	List: NUREG-0422, Requests Comments on Review of Encl marked-up Tech Spec Concerns Resulting from Differing Prof Opinion.Some Concerns Should Be Resolved on plant-specific Basis.Proposed Amend Requests & Reply Needed within 90 Days
References
RTR-NUREG-0422, RTR-NUREG-0737, RTR-NUREG-422, RTR-NUREG-737 NUDOCS 8507260195
Download: ML20128N380 (143)

Text

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f) > c(i g g UNITED STATES NUCLEAR REGULATORY COMMISSION 3* WASHINGTON, D. C. 20555

%, [ j -/l July 9, 1985

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Docket-Nos.150-3691 1 50-370-

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Mr. H. B. Tucker, Vice President Nuclear Production Department Duke Power Company 422 South Church Street Charlotte, North Carolina 28242

Subject:

Request for Comments on McGuire Technical Specification Concerns Resulting from Differing Professional Opinion In accordance with NRC internal procedures for expressing and handling differing professional opinions, a review of the McGuire 1 and 2 Technical Specifications has been conducted in response to concerns raised by a member of the NRC staff. The individual's concerns (Enclosure 1) result frem a l review of the Proof and Review copy of the Technical Specifications which i existed in mid-January 1983.

Our review of Enclosure 1 to data has determined tnat some of these concerns should be resolved on a plant specific basis. These are circled and numbered in the hand marked copy of Enclosure I and are further identified by indexing (Enclosure 2). Your comments on these plant-specific concerns are requested so that they may be considered in our further. review for final resolution.

Of particular interest would be your comments as to whether you believe a change to.McGuiro Technical Specifications is needed and, if not, your reasons tSereto. For those cases in which your coments reflect that a change to the McGuire Technical Specification is appropriate, a proposed amendment request, or your schedule for such requests, should be included.

Those items not identified in inclosure 1 and Enclosure 2 to be plant specific are either being considered by the NRC for generic resolution, have ,

been closed by NRC internal review, or are still under review. You may, of course, comment on any of these items should you wish to do so.

Your reply within 90 days of this letter would be consistent with our schedule for final esolution and, therefore, is requested. Contact our 8507260195 PDR 850709 P

ADOCK 05000369 >

PDR

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Mr. H. B. Tucker McGuire Nuclear Station Duke Power Company cc:

Mr. A. Carr Dr. John M. Barry Duke Power Company Department of Environmental Health P. O. Box 33189 Mecklenburg County 422 South Church Street 1200 Blythe Boulevard Charlotte, North Carolina 28242 Charlotte, North Carolina 28203 Mr. F. J. Twogood County Manager of Mecklenburo County Power Systems Division 720 East Fourth Street Westinghouse Electric Corp. Charlotte, North Carolina ?8202 P. O. Box 355 Pittsburgh, Pennsylvania 15230 Chairman, North Carolina Utilities Commission Mr. Robert Gill Dobbs Building Duke Power Company 430 North Salisbury Street Nuclear Production Department Raleigh, North Carolina 27602 P. O. Box 33189 Charlotte, North Carolina 2824? Mr. Dayne H. Rrown, Chief Radiation Protection Branch J. Michael McGarry, III, Esq. Division of Facility Services Bishoo, Liberman, Cook, Purcell Department of Human Resources and Reynolds P.O. Box 12200 1200 Seventeenth Street, N.W. Raleigh, North Carolina 27605 Washington, D. C. 20036 Mr. Wm. Orders Senior Resident Inspector c/o U.S. Nuclear Regulatory Commission Route 4, Box 529 Hunterville, North Carolina 28078 Regional Administrator U.S. Nuclear Regulatory Commission, Region II 101 Marietta Street, N.W., Suite 2900 Atlanta, Georgia 30323 ,

R. S. Howard Operating Plants Projects Regional Manager Westinghouse Electric Corporation - R&D 701 P. 0.~ Box 2728 Pittsburgh,' Pennsylvania 15230 L

o o tm.LU5UNt 1 MCGUIRE UNITS 1 & 2:

PROPOSED TECHNICAL SPECIFICATICNS CETAILED REVIEW OF " PROOF & REVIEV" CCPT I

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/, e SECTION 2.1 SAFETY LIMITS 2.1.1 REACTOR CORE The proposed T.S. requires that: "The combination of THERMAL POWER, pressuri:er pressure, and the highest operating loop coolant temperature (T,yg) shall not exceed the limits shown in Figures 2.1-1 and 2.1-2 for four and three loop operation, respectively.

APPLICA8ILITY: MODES 1 and 2.

ACTION:

Whenever the point defined by the combination of the nighest opg, rating, loop .

average temoerature and THERMAL POWER has exceeded the appropriate pressuri:er pressure line, be in HOT STANOBY sithin 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />, and comply with the requirements of Specification 6.7.1."

EVALUATION a) Concerning the title: SAFETY LIMITS / REACTOR CORE. Clarify if the numerical values in Figure 2.1 are meant to be Safety Limits, Limiting Safety Settings or Set Points.

b) Concerning Figs 2.1-1 What is the licensing basis for this type of re- ..

presentation, i.e. , RCS T,yg (*F) vs Fraction of Rated Thermal Power, anc the values in this figure. Reference 7, Figure 15.1.1-1, revision 7, is the existing licensing basis; it provices different ordinates, T,yg vs AT and includes descriptions of related acceptance criteria and limits wnich should also include boiling in the hot legs; it also provides direct links to the plant protection systems based on 2 out of 4 aT loco (individual) comoared with AT loop set point (indivicual), in the reactor protection system. Any such representation should also provida the basis for the SET-POINT metModology for each unit including values of all the carameters necessary to calculate OVERTEMPERATURE AT and OVERPOWER AT SET POINTS of related Table 2.2-1, REACTOR TRIP SYSTEM INSTRL' MENT TRIP SET POINTS; tnis ,

will ensure a complete set of Licensing Basis data against which tne pro-posed plant settings can be verified and amended as appropriate.

c) Representations of overpower protection (including reporting requirements) by neutron flux monitors on the Figure 2.1-1 are inapprcpriate. Neutron flux limits and related action statements are accressed under T.S. Sec-tion 3.4, [ Nuclear] Power Distribution Limits.

a) References to three loop oceration should be deleted as the plant is not licensed for such operation.

06/01/84 1 Revistor. A

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t e) Concerning description under Section 2.1.1 above. We propose this de-scription should clarify that the " combinations" presented are thos t allowed under " Anticipated Operational Occurrences" and not steady state conditions.

f) The FSAR does describe a constrained set of thermal hydraulic parameters for the Reactor Coolant Systam under steady state normal operating con-ditions upon which " plant safety" under Condition II, III and IV Occur-4 rences is established. These are generally described in reference 7, under Section 15.1.2, Table 15.1.2-2, and the programmed T,,g provided under reference 3, Figure 5.3.3-1; pressurizer pressure is provided under Table 5.1-1. (Related pressurizer level and steam generator levels will be discussed under T.S. Sections 3/4.4.3 and 3/4.4.5) Should not these values be included in'the Technical Specifications (in appropriate set point methodology) to meet the requirements of 10 CFR 50.36.

' For the thermal-hydraulic parameters represented in Section 2, the steacy state set points would be represented by a single line showing programmec Tavg against programmed AT for the given pressurizer pressure with pro-vision for a band of values to " allowable values". Appropriate action statements would be formulated providing a limited period of operation outside the range. Any changes proposed to such conditions need T.S.

amendments as they are part of the Licinsing Basis.

SUMMARY

The current method of representing Reactor Core Safety Limits is not clearly in accord with the Licensing Basis. Therefore it must be considered non- ~~

conservative and the Licensee shall evaluate and propose.

" REACTOR COOLANT SYSTEM PRESSURE .

2.1.2 The Reactor Coolant System pressure shall not exceed 2735 psig.

APPLICABILITY: MODES 1, 2, 3, 4, and 5.

ACTICN:

MODES 1 and 2 Whenever. the Reactor Coolant System pressure has exceeded 2735 psig, be in-

HOT STANOBY with the Reactor Coolant System pressure within its limit witnin 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />, and comply with the requirements of Specification 6.7.1.

MODES 3, 4 and 5 Whenever the Reactor Coolant System pressure has exceeded 2735 peig, reouce

, the Reactor Coolant System pressure to within its limit within 5 minutes, and

comply with the requirements of Specification 6.7.1."

I EVALUATION a) Is there not a need to forewarn the operator that as for 2.1.1, for normal

. steady state operation, the RCS pr.essurizer pressure shall not exceed tne 06/01/84 2 Revision A 4

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t values defined in Section 3/4.2.5 and 3/4.4.3. Safety evaluations for all ,

occurrences are predicated on those values and are invalidated if they are not sustained. If restoration cannot be achieved, there is a change frem the existing Licensing Basis and an appropriate request for a T.S. cnange would De necessary. ,

b) As for Section 2.1.1 above, is it not appropriata to clarify that the RCS Coolant System pressure shall not exceed [2735] psig under any Anticipated Operational Occurrence or Design Basis Accident.

I c) Where in the RCS system is the pressure limit to be observed og Reference 10, page 15.4-20, Revision 7 first para, shows that: "To obtain the maximum pressure in the primary side, conservatively high loop pressure drops are adoed to the calculated pressurizer pressure." What provision has been-made in the specified value or related instrumentation to conservatively account for this necessary correction. .

d) Pleaseclarifythatthevalueof2735psigisenactuaiSafetyLimit, being 110% of the Design Pressure of 2485 psig (reference 3, Table 5.2.2-2) and how is such a value determined by the operator when no set point, allowaole values and channel errors are provided for or defined.

e) Concerning Action Statement: MODES 1 & 2. This should consicer restora-

, tion of the RCS pressure to its required value for steacy state operation rather than within the 2735 psig limit.

Should MCCE 3 also be included in the action statement for MODES 1 & 2 as _

generally icentical concerns prevail except for the limited Appitcanility of Appendix G in T.S. Figs. 3.4-2.

f) Concerning MODES 3, 4 & 5.

How is the pressure limit of 2735 psig apolicable to MODES 4 and 5 nen recuced RCS temps. will cause consiceration of constrained Pressure /

Temperature limits (to Appendix G requirements] in T.S. Section 3/4.4.9.

Further, even MCDE 3 has an Appendix G limits of <2500 psig at RCS tamos.

of <350*F; reference T.S. Figs. 3.a-2.

. SUPMARY .

The current rearssentation of Safety 'imits for RCS pressure in this Sec-tion 2.1.2 is non-tonservertive with respect to the Licensing Basis. The Licensee shall evaluate and propose.

06/01/34 3 Re"ision A 1 =- 6., - e = ==. .,ny.,s m a

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3 TABLE 2.2-1. REACTOR TRIP INSTRUMENTATION SET POINTS 7

], [ Point Methodology, Table 3-4 and NOTE FOR TABLE 3-4 on page 1 l described as applicable to McGuire Unit 1, 50-359. At this date, the assump-tion has been made that this information also appifes to McGuire Unit 2, Occket I No. 50-370. Please docket this fact or otherwise provide the alternate j information.

L The writer finds the general approach to representing Trip Setpoints as 3,or g, a certain value is less than satisfactory; it is open-ended allowing overly conservative setpoints with unnecessary reactor trips. It appears that the Set-Point methodology may already have provided for expected errors in setting SETPOINTS so that this open-ended uncertainty is eliminated to a satisfactory

" manageable" quantity. The Licensee should clarify.

• j tem 3. Power Rate, Neutron Flux, High Positive Rate

]

Will a time constant of >2 seconds result in a slower response time, whien is less conservative.

]

./f, Item 4 Power Rate, Neutron Flux, High Negative Rate.

Will a time constant of >2 seconds result in a slower response time which is less conservative? . -

Reference 18 page 3-13, concerning Set Point Methodology advises that this "

value is not used in Safety Analyses. This appears in direct contradiction to reference 7, Section 15.2.3, page 15.2-12, revision 7, first para. The (Licenseeshallevaluateandpropose.

M Item 5: TS incomplete; should read as: Intermediate Range, [High] neutron flux.

[g[ Item 9: Pressurizar Pressure-Low The specified Trip Set:oint & Allowable values agree with those provicec un::er f set:oint methodology in reference 18. A disparity coes exist between :ne i related SAFETY ANALYSIS LIMITS given as used in Safety Analysis, i.e, 13a5 f psig in SETPOINT METHODOLOGY / Reference 18, Tacle 3-4, column 12 and the FSAR

• value for the same analysis in reference 7. Table 15.1.3-1 as 1835 psig. The -

Licensee shall icentify the correct value. (Nota also disparity with reference 7 " Analysis of Inaavertent Operation of EO:'S During Power Operation".

page 15.2-40, revision 43 item 7. " Reactor Trip ----- is initiated by low pressure at 1800 psia;" This is however relatively conservative with rescect to the other values used above.J t The Licensee shall review and clarify.

Item 17: The existing descriptor " Safety Injection Inout from ESF" should be replaced by " Reactor Trip from ESFAS."

06/01/84 4 Revision A

f e e The following items should be added, because they initiate Reactor Trip directly and independently of the SI signal.

17a) Pressurizer - Low Pressure (Safety injection)

The additional qualifier (SI) is generally used to distinguish this from item 5, Reactor Trip on Pressurizar Pressure-Low 17b) Containment Pressure-High 17c) Low Steam Line Pressure (subject to P-11 block) 17a) Hanual Safety Injection Item 12: Low Reactor Coolant Flow

, a: Concerning Reactor Trip on " Low-Reactor Coolant F)ow in One Loop."- .

Reference 7, Section 15.2.5.1 states that "Above approximately 50% power, Permissive P8 allows low flow in any one loop to actuate a reactor trip."

Please explain why there is no anticipatory signal for this circumstance fe under frequency, undervoltage, loss of RCP breaker. Such anticipatory sig.nals are provided below P-8 when. safety consequences are more conservative for this facility. (See later 12b.)~ Is this adequate conformance to diversify require-ments of Criterion 22 - Protection system independence.

b. Concerning Reactor Trip on " Low Reactor Coolant Flow "In Two Loops Below P-8.

The plant is not licensed for operatien with only 3 loops operating in MCOES 1 and 2 below P-8. Please explain why you therefore propose a trip cased on Loss of Flow in 2 locos instead of only one, at these conditions and wnich is not in conformance with GDC 20, " Protection System Functions." Information is provicea under reference 7, Section 15.3.4.1 to show that Acceptance Criteria would not be exceeded but as indicated above it is outside the current licensing oasis and should theref' ore be excluded.

This licensee should evaluate our concerns in items 12a and 125 acove in sa conjunction with those of item 18.b.a of this same review of Tacle 2.2-1, and procose. This can be interpreted as a generic issue.

lj Item 13: Concerning Steam Generator Level-Low, Low h Reference 18, page 3-13 Note 12 describes the Safety Analysis L!mit for tnis item as the value in Table 2.2-1 of the W STS olus 10%. For conservatism, should the Safety Analysis Lis.t be the W STS value less 10%; is this neces-sarily conservative for all Licensing Basis occurrences?

Item 14: When two or more RCP circuit breakers open, above Permissive 7 (1C*

oower), Reactor Trip deriving from undervoltage of the Reactor Ocolant Puros is also initiated, reference 7 Section 15.2.5.1 and reference 5 figure 7.2.1-1 06/01/84 5 Revision A

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note 4. It is proposed that a notation to this effect should appear under this item.

Item 21 (Peccesedh [ Reactor Trip on) Reactor Coolant Pump Breaker Position Proposed:

In accordance with the Licensing Basis FSAR, indicating that opening of two or more circuit breakers actuates the corresponding undervaltage trip

. relay above Permissive 7 (10% power); reference 7, section 15.2.5,1.

Item 18b: Low Power Reactor Trips Block, P-7 a) This T.S. provides that when power level is less then Permissive P7 (with P10 (Nuclear) or P13 (turbine) powers of less than 10%) the undervoltage (and RCP breaker position), under frequency and low flow reactor trips are blocked and will allow the reactor to remain untripped, and therefore at 10% power,. on loss of offsite power. ,. -

The F5AR in reference 5, item 7.2.2.1.2d which describes this permissive provides no safety evaluacion of the consequences. Accident Analysis in Reference 7, section 15.2.9 for " Loss of Offsite Power to the Station '

Auxiliaries" is based on protection provided by these trips which are now blocked, and no evaluation is provided to show an acceptante RCS response under these particular circumstance. The existing FSAR, reference 7, Section 15.2.9.2 and related Table 15.2.9-1 shows acceptante natural circulation, but at a maximum power level of only 5%.

Accident Analysis in Reference 7, Section 15.3.4 " Complete Loss of Forced Reactor Coolant Flow" also depends on this protection, and no evaluation is provided to show an acceptable response by the RCS system from the P-7 power levels. This also applies to Section 15.4.4, " Single Reactor. Coolant Pump Locked Rotor."

There are accitional events'potentially arising from this item which have not been analyzed. These include a circumstance in which a normal turcine load rejection from just below the P-8 power level could result in a sequence in which power to RCPs are lost after both Nuclear and Turcine Pcwer signals are reduced below 10% (P-7) so that reactor trip on this loss of power event coula not occur, but with residual core heat fluxes at substantially greater than 10%

in the early phase of the event followed by a 10% steady power level [ Note also, i

tnat below P-7, a number of other reactor trips are also blocked including Pres-suri:er Water Level-High, Pressuriter Pressure-Low and Pressuri:er Pressure-High) ,.

The situation is one in which Condition II, III and IV occurrences are not protected in accordance with GDC 20, Protection System Functions: "The r

protection system shall be designed (1) to initiate automatically the operation

! of appropriata systems including the reactivity control systems, to assure that specified acceptacle fuel design limits are not exceeded .J a result of anticipated operational occurrences." It also introduces an additional occur-rence, i.e., a failure to automatically trip the reactor, on top of the initial occurrence, and which in itself, and in comnination with the initiating eccur-rence has not been evaluated.

It nas not been Regulatory Practice to allow a Condition II occurrence to ce followed by a Condition III or IV occurrence in the course of protective actions.

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The if censee should evaluate the restoration of reactor trip on " low flow" trips a down to and including MODE 2 (MODES 3-5 are discussed later) to be in conformance i with G.O.C. 20 " Protection System Functions," and propose. As part of this evaluation, the Licensee should verify performance under these T.S. conditions and review for, and evaluate, Licensing Basis occurrences affected by this T.S.

requirement to show that all Regulatory Acceptance Criteria for Abnormal Operating Occurrences and Postulated Accidents are currently satisfied, making appropriate allowances for any manual Operator Action required. These events I should include Loss of Off-Site Power to the Station Auxiliaries, Complete .

Loss of Forced Reactor Coolant Flow and Single Reactor Coolant Pump locked Rotor. (It should be noted that other reactor trips such as Pressurizer Water i Level-High and Pressuri:er Pressure - Low are also blocked under these condi- i tions. Steam Generator Water Level-Low Low remains available together with  :

Auto-initiation of AFW pumps. Steam Generator High High Turbine Trip is avail- t

, a:1e, but does not trip the Reactor at these low power conditions .(below P-8).]

I Until the required re-evaluation is' completed, the' proposed T.S. must ce considered non-conservative in respect to Regulatory Requirements. Adcitionally it can be interpreted as a Generic Issue.

b) The current description of this Functional Unit is incorrect. It is not

" Lower Power Reactor Trips Block P-7." It is: "High Power Reactor Trips Block," by absence of Permissive P-7 and occurs when:  ;

1) P-10 is less than the Trip Set Point and

2) P-13 is less then the Trip Set Point c) This' TS orovices that when power level is less than Permissive P7 (with t P10 -(Nuclear) or P13 (Turbine) powers of less than 10%), reactor trip' on Pressuri:er Pressure-Low and Pressuri:er Water Level-High are both blocked.

c(1) Concerning Block of Pressuri:er Pressure Low - Reactor Trip:

The FSAR in reference 5, item 7.2.1.1.2.0.1 states that this trio is not recuired at low power levels.

The pressuri:er pressure low - reactor trips are used as both primary anc sacx l up in a numoer of Condition II Condition III anc Condition IV cccurrencar, all involving breaks in the primary and secondary systems, reference 7, tacle 7.2.1-4 ,

(3 of 5). Although safety injection is subsequently emotoyed in almost all g these situations, earlier reactor trip on pressuri:er pressure low - is depended upon instead of the later reactor trip on pressuri:er pressure low - (Safety Injection). The worst situation for .most of these accicents is that of maximum power level referacce 7, Table 15.1.2-2. No evaluations are provided for :ero power level.

It is possible for tnese breaks in tne primary and secondarf systems to occur at less than 10% power level down to and including the startup condition (with 4 9CS toops running) ie MCOES 1 & 2. (Suen breaks in MCOES 3-5 are discussed later). With the proposed TS, reactor trips for tnese breaks would be celaved to be initiateo later by the ESFAS.(SI) related signals. The licensee snouic '

provice a safety evalution of these circumstances and which is not based i. con .

arguments relating to procacility of the events. The evaluation snould pr:vice 06/01/84 7 Revisien A

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for the event to occur immediately subsequent to any normal operating transient providing the most conservative set of conditions prior to the event such as a complete load. rejection using steam dumps from the P-8 level.

Untti there has been a re-evaluation of these circumstances, the proposed T.S.

must be considered non-conservative in rescect to Regulatory Requirements.

Accitionally it can be interpreted as a Generic Issue.

I

[g Accidental Depressurization of the main steam system is from Zero load. It is unclear from reference 5 Table 7.2.1-4 (5 of 5) if for this event, reactor trip on Pressurizer Low Pressure is expected to occur before Safety Injection (when it would not be available at zero power) or whether it is expected to occur from the pressurizer pressure low - (Safety Injection) signal if it initiates S.'I.,

or,from S.I. initiated by other initiators. The Licensee shall clarify, and hence its validity with resoect to the absence of the signal caused by P7.

j cii) Con:erning Block of Pre'ssur'izer Water Level-High Trip .

This pressurizer water level-high trip is a principal element of the Overpres-sure Protection System for W PWRs as fully discussed in Topical Report to.

reference 27.

Amongst Licensing Basis events, this trip is used as primary or back up on Uncontrolled Rod Cluster Control Assemoly at Power. Uncontrolled withdrawal from a subcritical condition (at below P10) is protected primarily by other trips.

Among Licensing Basis events this trip is also used on Loss of External electric load and/or Turoine Trip. Most severe design basis consequences are from full power. Such an event at less than the 10% Set Point (P-10 & P13] is within the normal control range of the reactor (without steam dumo) with the expectancy of no values exceeding normal control band (and thereby not approaching T.S. LimitsJ.

The blockage of these trips is consistant with the Design Basis Events and ex-pected behavior of the Control System. However this does not address the fact that Design Basis events only define the outer envelcoe of expected severity wnich is excec*ed to cover a large number of less severe occurrences, uncefined.

It acpears singularly inaccropriate to remove these protection cavices *nicn could play a primary or backuo role in such circumstances. For examole, refer-ence 5, page 72-27 item 7.2.2.3.4, " Pressurizer Water Level " descrites the role i

of the Pressure Water Level trip in preventing liquid Coolant discharge througn ,

the safety valves during a failure of the Pressurizer Water Level (P%L) controller at full pc,wer. Failure of PWL controller could fill the pressurizer witnin i

s hour or longer, but T.S. Table 4.3-1 shows a channel check on only a snift basis. Further, a single channel failure to low could cause overfill of the pressurizer (through the level control system) and with subsequent permissable failure of a second channel could remove the alarm expected fron. 2 out of 3 to t

that no alert is given the operator which would be contrary to the recuirement l of the FSAR.

There is no discussion on the imcortance of its use at low powers althougn the genersi System Description provided under Section 7.2.1.1 and its l

06/01/84 - 8 Revision A i

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protective actions is no less apprcpriate at 0-10% power, as it is at "

higher power levels.

It is proposed, reference 5 page 7.2-6 that Pressuri:er Water Level-Hign Trip below P-7 is automatically blocked to permit start up. Whereas this is under-standable in MODES 6, 5 and part of 4, it is not a valid proposition once a bubble is formed in the pressuri:er in MODE 4 and the Pressurizer Level Control can be placed in AUTO. Considering the attention required of all other manual a actions during MODES 4 through 2, it is not appropriata to. remove the automatic

. protection of the RCS boundary. Further, in MODES 4 and 3 it could be one of the only effective trips available because of the potential non-viability of -

Pressuri:er Pressure High and non-applicability of existing Pressurizer  ?

Pressure-Low.

The Licenee should evaluate the impact on safety by blocking the Pressure Water Level-High trip below P-7, including all the concerns discussed above. j

. This item can be interpreted as a generic issue. This'could be considerec non- g conservative in respect to Regulatory Requirements because of the absence of f automatic protection in accordance with 10 CFR 50, GDC 20 " Protection System Functions," both for reactivity control systems, and overpressure protection systsms.

c(iii) The absence of permissive P-7 (on P-10 and P-13] introduces new events to evaluate for safety. This requires related Safety Analyses Limits and  !

the Licensee shall advise what these are for each of P-10 and P-13 and how these are combined for 7-7. ..

l Item 18(f). Preposed new item: High Power Reactor Trip on Turbine Trip; Block by aosence of P-8.

The Anticipatory Reactor Trip on Turbine Trip required by TMI Action Plan II.K.3.12, 1.s bypassed below P-8. The SER is provided in reference 15, Item II.A.3.12, and reference 21 for McGuire Unit 1. We have issues no related final SER for McGuire 2 at this time. Note the related Sasis vili need to be amended.

Item: Loss of "PCWER" Their is a need tp prescribe the conditions under wnich a reactor would ,.

trip directly from a " Loss of Powerd condition other than tnose deriving from other Functional Units. This is a substantial emission from the Tecn-nical Soecifications.

Item: General - This is a need to identify potential blockage of eacn of these Reactor Trip Functions by Plant Logic and any related manual action, e.g. ,

3 P-7, f P-11 with manual blocks-9 etc. This enaoles improved perception of real levels of engineered protection than is currently available. Table 3.3 1 contains only approximate information concerning plant situations at wnien protec-ion levels are enanged. It also contains NON-CPERABILITY MODES wnich are not pre-determined by Plant Logic.

I 06/01/84 9 Revision 4

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1 SECTION 3.4.1 REACTIVITY CONTROL SYSTEMS Section 3/4.1.1 00 RATION CONTROL / APPLICABLE MCDES 1. 2*, 3 a'nd 4 T.S. Pages 3/4 1-1, 2, 2a: Reference 16; page Q 212-47e states " Operating Instructions require that boron concentration be increased to at least the cold shutdown boren concentration before cooldown is initiated. This requirement insures a minimum of 1% celta k/k shutdown margin at an RCS temperature of 200*F." This is used as a means of protecting against NON-LOCA Accidents during startup and shutdown.

Since this proposal to increase boron concentration is a limiting condition for operation required for safe operation of the facility from and including MCDE 3 down to and including MODE 5, please advise why this does not aapear in the Technica.l Specifications in ,accordance with 10 CFR 50,.36(c)(2).

T.S. Page 3/4 1-1 and 2 specifying a shutdown margin of 1.6% delta X/< over '

MODES 1 through 4 should be modified to exclude MODES 3 and 4, and SHUTDC'nN I MARGIN T4yg should be changed from >2004 to 1557*. -

A new T.S. Page 3/4 1-2(a) should be added for BORATICN CONTROL SYSTEMS in MCOES3through5.fromTlEE11beincreasedtoavaluewhichwillgivea<

concentration in the RCS 557'F throu i shutdown margin of 1% delta K/K at 200*F.

Safety Signficance: These actions are necessary to bring the safety status of the plant into conformance with the Licensing Basis. Without this, the ,~

l plant is in a less than conservative MODE which has not been evaluated.  !

Further, it appears that OPERA 8ILITY REQUIREMENTS of Table 3.3-1, REACTOR TRIP SYSTEM INSTRUMENTATION and TA8LE 3.3-3 ESFAS INSTRUMENTATION may be conditioned on these higher Baron Concentrations so that ommission of Additional Baron Concentration in accordance with Reference 16, page Q-212-47e makes for an inconsistent and nonconservative level of protection for all NON-LOCA events for T,yg 1 557*F.

The proposed T.S. nignt be accentable if all events were analyzed in MCOES 3 tnrougn 5 and the CPERABILITY REQUIREMENTS OF TABLIS 3.3-1 and 3.3-3 reviewed.

Reference 11, page 15-2, first para. precludes any baron dilution after a reactor scram until the neutron flux level is below the level of the source range high flux level alarm. This is effectively an LCO that is not included "

in the proposed T.S.

The proposed T.S is non-conservative with respect to the Licensing Bases.

The Licensee shall evaluate our concerns under this Section 3/4.1.' and propose.

I

{ f TS pace 3/4 1-6. MIN' MUM TEMPERATURE FOR CRITICALITY l the existing minimum temperature for criticiality (in MODES 1 and 2) is given I as 551*F. Please advise wny this value is less than the programmed set point minimum value of 557'F in reference 20, fig. 5.3.3-1. Accident evaluations g

for events from zero power are predicated upon this set point of 557*, and any )

06/01/84 10 Revision A i

2 kariation tnerefrem in either direction would be unacceptable Reference our m mments unaer section 2.1.1.r.

An example of a safety imoact is for the Design Basis Main Steam Line Break Event which is initisted from zero power in MCOE 2 from a set Point Tmin of 557'F. Any " increase" in this value (at given shutdown margin) would lead to conditions less conservative than the design basis.

To be within the Licensing Basis, this T5 Section 3.1.1.4 should therefore provide that the Temperature for criticality (at zero power] shall be a set

. point value of 557'F with appropriate surveillance requirements. The Appli-cability is for M00E5 1 and 2.

The proposed T.S. is non-conservative with respect to the Licensing Basis. The Licensee shall evaluate, including the above concerns, and propose.

', Section 3/4.1.2 BORATION SYSTEMS ,

T.S. Page 3/4 1-7: Concerning "BORATION SYSTEM, FLCW PATH - SHUTOOWN.

APPLICA8LE M00E5 5 and 6:

The current T.S. requires an (unidentified) charging pump to supply Boron to the RCS. Current Licensing constraints on ECCS operation discussed under Section 3/4.5 Emergency core cooling systems" require that only one centrifugal l.

charging pump is permitted to be in operation from a condition of 1000 psig/425*F in MODE 3 down to RHR operation commencing with MODE 4. In MCDE 4, a similar and parallel requirement for overoressure protection in the RHR mode with -

water solid operation extends this requirement througn MODE 4 to MODE 5; reference 11. page 5-1 where it is described that under RHR operation, the "only remaining centrifugal charging pump could cause an overpressure transient as a result of inadvertent start" but that "The Licensee has shown that [in this case] the 10 CFR 50 Appendte G Limit is not reached.

Charging oumo requirements in McCE 6 are defined by reference 10, Sec-tion 15.2.4.2, item 3 under " Dilution During Refueling" in anien a pre-condition for the " uncontrolled !cron 011ution Event" is that "tne enarging purcs are inocerative."

These circumstance permit only one charlino oumo, which must ce a centrifugst pump only, in operation from 'stanoey tat 1000 psig/425'F) througn to MCOE 5",

therefore the term SHUT 00hN in the title and the APPLICABLE MCOES 5 and 6 g should ce replaced by these conditions. Also, the descriction of the enargi9g .

pumo should te expanded by the term " centrifugal" together with the proviso that "tnis centrifugal enarging pump slso be the same and only pump allowed for ECCS and other acerations under these circumstances."

The procosed T.S. is non-conservative in resoect of the Licensing Basis. The Licensee shall evaluate and propose.

06/01/g4 11 Revisice ,'.

q T.S. Pace 3/4 1-8. Concernino: " FLOW PATHS - OPERATING" in APPLICA8LE MODES 1 Z. 3 and a.

The Licensing 8 asis ECCS requirements. discussed under Section 3/4.5 EMERGENCY CORE COOLING S'fSTEMS of this report do not constrain charging pumo operation noove 1000 psig/425'F. Therefore the existing provisions on this T.S. page for charging pumps remain valid with the exception that APPLICA8LE MODE 4 should be deleted and MODE 3 must be conditioned as M00E 3 (Down to .

1000 psig/425'F). Further the title should be changed to incorporate these constraints.

i The proposed T.S. is non-conservative in respect of the Licensing Basis. The .

Licensee shall evaluate and propose.

The ACTION statement should be revised to be consistent with the Scration Requirements adopted out of item "Section 3/4.1.1" of t.his report. ,

T.5. Pace 3/4 1-9 concernino: CHARGING PUMP-SHUTOChN Consistent with the work of the previous TS Section 3/41-7 of this report, this title should be changed to: CHARGING PUMP "Stanchye (at 1000 psig/

a25'F) througn to MODE 5. Additionally, under subsection 3.1.2.3 modify to only Lne centrifuaal charging pump shall be OPERA 8LE. APPLICA4ILITY is changed from W E3 5 and e to MODE 3 (at 4 1000 psig/425'F), 4 and 5. M0.0E 6 is deleted.

Surveillance Recuirements under subsection 4.1.2.3.2 must reflect the require- *-

i monts of later SECTION 3/4.5 ECCS of this report fn which "All centrifugal,  !

(and reciprocating) charging pumps excluding the required CPERA8LE pump shall '

be demonstrated inoperable by" addi11onal features to those already described in  !

this subsection, namely, "by verify'ng that the motor circuit breakers are  !

secured in the open position by beina onened. locked and taoced; the alternate of isolation from the Reactor coolant system ey at least two isolation valves 1 with breakers for the valve operators beina onen. locked and tacoed has not been provided.' (reference 12, page 6-6 concerning racning anc locxing out of pumos; also reference 11, pages Q212-47 and 47a)

The proposed T.S. is non-conservative with respect to the Licensing 3 asis. The Licensee sna11 evaluate and propose.

T.S. Psae 3/4 1-10 Concernina: CHARGING PUMPS - OPERATING AND AP8'.!CABILI'/ -

MUDE5 1. 2. 3 anc 4 l

This is directly related to the proposed changes under Item T.S. Page 3/4 1-8  !

of this report. Consistent with that discussion, the title should be changed to delete M00E 4, and MODE 3 conditioned to (down to 1000 psie/425'F) '

Item 4.1.2.a.2 under SURVEI' LANCE

. REQUIREMENTS does not now acoly since it refers to conditions 1 300*F which are not now covered by this section, being f limited to a minimum of 1000 psig/425'F in M00E 3. The same comment applies to footnote (,,,, concerning one only centrifugal charging pumo at f, 300*F.  ;

The proposed T.S. is non conservative with respect to the Licensing Basis. The  !

Licensee shall evaluate and propose l

06/01/84 12 Revision A j i

d

T.S. Pace 3/4 1-11 Concernino: BORATED WATER SOURCE - SHUTDOWN .

This title (and related Applicability MODES 5 and 6) should be changed to SCRATED WATER SCURCE - MCDE 3 (1000 psig/425'F) THROUGH TO MCOE 5 to te compatible with the changed title to T5 pages. 3/4 1-7 and 3/4 1-9 discussed earlier since this page refers to borated watar sources for situations there ,

described. 1 E

Additionally, (by letter to reference 17] the Licensee has committed to provide and T.S. an operable level detection system with a specified " minimum level".

This has not been included in the T.S. and it is proposed that it form the subject of an additional item 3.1.2.5.a.4). Surveillance requirements should .

y t

be included under 4.1.2.5.a.4) in which the borated water source would be demon-strated OPERABLE by verifying minimium levels in the system.

Further, an additional surveillance should verify the availability ofc Level-Detection-(2 indicators / tank) and related high, low and low-low level alarms. l Clarify whether the LCD values proposed are Safety Analysis Limits or Set Point Values.

An appropriate modification may need to be made to the Baron Concentrations and volumetric recuirements in the Boric Acid Storage System in these MODES 3 (1000 psig/425') through 5 to provide for the increased Baron Concentrations required from the Licensing Basis in these MODES discussed in this report uncer i T.S. page 3/41-1, 2 and 2a. ..

Why is the refueling water storage in MODE 5 proposed as only 26,000 gallons wnen reference 6, page Q212-57, revisi'on 25, under Case-3 provides that in ,

MODE 5, in the event of loss of cooling by a fait closed RHR/RCS isolation valve the charging pumo could provide feed and bleed cooling througn the PORvs for up to 5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> from the RWST and subsecuently the RHR pump and heat exchanger would re-circulate and cool from the containment sumo. Would not this recuire an unchanged recuirement from MODES 1 througn 4 of at least 372,100 gallons.

The proposed T.5 is non-conservative in respect to the Licensing Sasis. The Licensee sna11 evaluate, including all our concer9s above under T.S. 8 age 3/4 *.-11.

and propose.

T.S. Pace 3/4 1-12 concerninc: 80 RATED WATER SCURCES - OPERATING (in relstad Aeolicaone MCOEs 1. 2, 3 ano 4)

BORATE 3 This title, and related applicability modes, should be changed to:

WATER SOURCES - MCOES 1, 2, and 3 (Down to 1000 psig/a25'F) to be compatible with the changed title to T.S. Pages 3/4 1-9 and 3/4 1-10 discussed earlier, since this page refers to corated water sources for the situations there described.

Additionally (by letter to reference 17] the Licensee did commit to provice and T.S. an operable level detection system with a specified minimum level. This has not been included in the T.S. and it is proposed that it form the subject Additional surveillance requirements of an additional item 3.1.2.6.a.a).

snould me included under 4.1.2.6.4.4) in whien the borated water source .cuid os demonstrated CPERABLE by verifying minimum levels in the system.

1 13 3evision 4 06/01/34

Further, an additional surveillance should verify the availability of Level Detection (2 indicators / tank) and related high, low and low-low level alarms.

Clarify wnether the LCO values given are Safety Analysis Limits or Set Point Limits.

An appropriate modification may need to be made to the Soron Concentrations and volumetric requirements in the Soric Acid Storage System in MODE 3 down to 1000 psig/425'F to provide for the increased Boron Concentrations required from the Licensing Basis in this MODE discussed in this report under 75 page 3/4 1-1, 2 and 2a.

The Aasence of required LCOs makes the proposed T.S. less conservative than the Licensing Basis. The Licensee shall evaluate, including our concerns unoer TS Pages 3/4 1-12, and propose.

'T.S Pace 3/4 1-13a. Paccosed concerninc INSTRUWENTATION IN 90CES 3. 4, 5 anc o SER Supp 1, reference 11 page 15-2 reouires a Technical Specification that "During startup and shutdown, the app 1tcant will rely on the source range high flux alarms to alert the operator that a dilution event is occurring. This assessment is based on setting the alarm at a l'svel of 5 times the background level. The licensee is to maintain the source range alarm setpoint at this level or lower any time the plant is in the cold shutdown Mode. The set point is to be checked and adjusted on a weekly basis if in the cold shutdown 4

mode for an extended period." *'

' This SER requirement has not been provided in the Technical Specifications. .

Please discuss provision under a proposed new item under Section 3/4.1 REACTIVITY CONTROL SYSTEMS, entitled " INSTRUMENTATION" in which these require-ments would be proposed for Applicable MODES 3, 4, 5 and 6.

A similar provision is provided under Refueling, TS page 3/4 9-2 INSTRUMENTATION and is acolicaole only to MODE 6. Since it is a part of " Reactivity Control Systems" and accifesble over additional MCDES, it snould be previoed in this context also as discussed above.

The proposed T.S. is less conservative than the Licensing Basis. The Licensee shall evaluate and propose. -

T.S. Pace 3/4 1-20 Concerninc: SHUTOCWN R00 INSERTION LIMITS T.S. Paos 3/4 1-21 Concernino: CONTROL R00 INSERTION LIMITS a) Specifications for limiting conditions of operation on the positions of these movable control assemblies apply only to M00ES 1 1 2. There is no Tecnnical specification on positions in MODES 3 5 althougn T.S. Page 3/4 1-18 concerning " Position Indication system - shutdown" requires operability of a Rod Position indication system in MCOES 3 througn 5 when the reactor trip system breakers are in the closed position.

06/01/84 14 Revision A

O .

1 It is proposed that in general, Technical specifications are required by 10 CFR .

e 30:46 to be placed on the limits of movable control assemolies in these modes to limit the consequences of Condition II, III and IV events which may occur, l.

unless analyses and evaluations shew that these are unnecessary.

An example of the need is reflected in the memo to reference 26 in which rod positions for Baron Dilution events are specified from Refueling through to Hot stanoby as All Rods Out (Mode 6, Refueling) and, All Rods In with Most

• I Reactive Rod Stuck Out, for Hot Standby through Cold shutdown. Further, applicants may opt to assume a more limiting initial control rod position -

which would however need to be justified. ,

I The Baron Oilution event for McGuire has "apparently been* made acceptacle by .

proceoures requiring the RCS to be filled with Borated (approx 2000 ppm) .

f water from the refueling water storage tank prior to " Start Oo"; reference 7,

. page 15.2-15, revision 10. Reference earlier discussion on TS. Pages 3/4 1-1, 2,and 2 a. This is an LCO and should appear in the proposed T.St .

With the existing T.S. without the required increase in Soron concentration, there is no guarantee that a return to power during dilution will not infringe current RCS Safety Criteria. Under those circumstances a T.S. on the Position at shutdown of Control Rods is required unless an acceptante safety evaluation is suomitted to show the contrary.

. . t In general, also, the same concern applies to any other Condition II, !!! and IV occurrence wnich can lead to a return to oower in these Modes. Until these circumstances can be shown to result in acceptacle consequences without a T.S. -

on the position of these movable rods, then 10 CFR 30:46 would require such a Technical specification. In this evaluation, cognizance also needs to be given to the reduced operability requirements for all Reactor Trip Instrumen-tation and Engineered Safety Features Actuation Instrumentation in these MCOES (3 through 5). This is particularly significant with the proposed I.S.

on Boration Control where resulting shutdown margins are substantially less than these provided by the curesnt Licensing Basis.

The Licensee shall provide analyses and related safety evaluations to justify his current absence of Technical Scecifications in respect of $HUTCC'aN anc CCNTRCL RCO positions during McCES 3 througn 5. Without this, tne procosec T.S are non-conservative with respect to the Licensing Basis, b) Overpower (AT) and overtemocrature (AT) protection systems incorcorate automatic limits (Rod stops) on control rod insertion to maintain Safety Analysis Limits on " Power Distribution" in the Reactor Core curing power runcack.

Please advise wny there are no surveillance limits and requirements for these Rod stops in your Technical Specifications to meet the requirements of 10 CFR 50.36. Without these, the croposec T.S. must be considered non-conservative.

06/C1/34 15 Revision 4

I ,,

I.

Section 3/4.2 POWER DISTRIBUTION LIMITS .

t -

Sectien 3/4.2.1 THROUGH 3/4.2.4 POWER DISTRIBUTION LIMITS ASS has not reviewed these sections en the understanding that they are the primary responsibility of Core Performance Branch.

l Section 3/4.2.5 DNA pAAAMETERS AND TABLE 3.2-1 DNA PARAMETERS The current information does not adequately represent all these perameters necessary te ensure "acceptele" RCS operations, including DNS, under all .

Licensing 8 asis Conditions II, !!! and IV.' .

The necessary parameters are disgussed and described under Section 2.1.2.

Reactor Core, item f, of this report. If they are logically recrosentee under 2.1.1. (and elsewnere), wny are they also reoresented here?

Evaluation i a) DNS presenta only one Acceptance criteria for acceptete operation of the ACS: There are others including fuel element clad failure and Appendix 4 requirements depending upon the occurrence being considered. Additionally there are RC5 overpressure, steam generster everpressure and Hot Leg Boiling Criteria. .

As indicated in our comment in Section 2.1.1, item f initial conditions which coveralargerN*efvariablesthanthesepresentedInTable3.2.1,incomeina-tion, determine RCS safety in the necessarily broadest sense.

It is suggested that this section be deleted, and the relevant information be supp11ed under T.S. Sections 2.1.1 where it belongs and where it has been discussed. ,

D) Concerning Table 3.2-1. The value for Reactor Coolant System T qiven as 1593'F is not in acesroance with the F5A4, reference 3, Figure l'U31 wnere a value of 548.1'F f a given as the programmed T for AATED THERMAL PC'atA Conditions. PleaseexplainthedifferenceandIMiainwnysetnointanc alloweele values should not ne provided. As a Setooint, the procosed 75 value is non-conservative with respect to the Licensing Sasis. '

Please emplain why a related pewer level has not been ascribed to this temperature.

Pleaseemplainwnyprogrammed7)8 has not been given for sere powl operationof 557.0*F (also reference (Reference again our3.Section Ff gure2.1.1 5.3.31 item f).

c) Concerning Tatte 3.2-1 Pressurizer Pressure. Please explain the basis for the given value of 3 2230 psia wnen Infomation in reference 20. Taele 4.11 (1 of 3) shows a "5ystem Pressure, Nominal" of 2250 psia and Section 15.1.2.2 Taele 15.1.2 2 makes provision for a total of 30 psi for steady state fluctu-ations and measurement error. Have you quote # a Setpoint value, or an allowaele 06/01/84 16 Revision A 2

I e .

i value; both should be available. As a Set;oint, the proocsed 7.5. value is non- **

conservative with respect to the Licensing Sasis for CNBR, and conservative for overpressure protection.

d) 'ehy should not programmed T,yg be provided under T.S. Section 2.1.1 e) Why should not Pressurize Pressurer be included both under T.S. Section 2.1-1 and T.S. Section 3/4.4.3 Pressurizer, a l

f) As discussed in Section 2.1.1, subsection f, additional parameters necessary to the validity of Accident Analyses in Section 15 include Pressurizer Level .

(See our review under Section 3.4.4.3. T.S. Page 3/4 4-9) and Steam Generator P Levels under Section 3/4.4.5 T.S. Page 3/4 4-11).

I CONCLUSICN

+ The parameters proposed by tne T.S. as "CN8R PARAMETER" under TA8LE 3.'2-1 are an I incomotete set and inaceouately defined in terms of Set Points, Allowsole Values and Safety Analysis limits. All this necessary information is availaole from the existing Licensing Basis and their incomplete and inadeguate recre-sentation crsates a non-conservative situation with respect to the Licensing Basis. The Licensee shall evaluate and propose. This is only partly a generic ,l proolos arising from an inadequate representation in the M 575.

i r

06/01/84 17 Revision A

TABLE 3.3-1 REACTOR TRIP SYSTEM INSTRUMENTATION i

T.S. Pace 3/4 3-2.

Item Sc: Source Range, Neutron Flux Does this channel provide an alarm only function, or an alarm plus trip j function. i b[OuringshutdowninMODES3,4and5,withreactortripsystembreakersooen.

f Source Range, Neutron Flux, channel opera 0111ty requirements specify only one

cnannel operable, and if this same channel is seing used to meet the Baron l

) dilution alarm requirements of proposed T.S. Page 3/4 1-13 (a), taen it is not in accordance with the Soron 011ution Requirements of the FSAR for whien at i

least 2 operaale channels would be required; reference 8, page Q212-24, item 212.58. The Licensee snell evaluate and precose. Currently, this- -

accears non-conservative.

Item 6a: This Tecnnical Specification concerning Operability of the source Range Neutron Flux is unclear. It species operacility of the Source Range i l

' Neutron Flux trip Below the P-6 (intensediate Range Neutron Flux Setcoint) during startup in MCDE 2; the Licensee shall advise if this " start up" enannet i

is reoutred to be Opersale to get Reactor trip in MCOES 3, 4 and 5.

• Items 1 througn 5: The FSAR, Reference 5. Table 7.2.141 of 5 shows the i a

Power-Range Neutron Flux Trip Low 5etpoint and Hign Setooint, and the l Intermediate Range Hign Neutron Flux Trip, and the Source Sange Hign Neutron Flux Trip, all being used on events being initiated'from a "suscritical"  !

condition. However, Tacle 3.3-1 shows that except for the Source Range Neutron Flux items 6b and 6c, all the Trips are inoperable in the suberitical 4

M00E5 3 througn 5. Furtner, there is a note d) in the column entitled Tecn.

t Spec (c) of Table 7.2.1 4 wnien states that "A technical specification is not recuired (for the Intermediate Range High Neutron Flux Trip and Source Range l Hign Neutron Flux Trip) tecause the trio function is not assumed to function ,

. in Accident Analyses. Please note furtner that this cosition is follo ed i througn in Tacle 3.3-2 Items 5 and 6 in that a resocnse time is not prov14ec for tne Intermediate ano Source Range Neutrcn Flux trips, because it is cro-  !

sosed as NA (Not Apolicaele). Please evaluate tne accarent parsdox that tne Source Range Trip is the only nuclear Flux trip recuired to be CP!RA8LE in tne suceritical MCOES 3 tnrougn 5, and yet there is no Tecn Saec proposed for it. .-

At this moment, aesence of CPERA8:LITY reouirements for the Power Range Neutron Flux Trip, Low 5etootnt, in MCOE5 3 througn 5 would acosar to constitute a l disparity with the Licensing Basis FSAR and in a less than conservative manner.

The Licensee shall evaluate and propose, those safety related neutron Flux trips which =culd De appropriate to use and available to trip the reacter for any of those events causing a return to power and under circumstance in whPn a safety injection initiator is not availaolo, during MODE 5 3, 4 and 5; and provide tne 1

related Set Points, Allowante values and Safety Analysis Limits. Alternately,  ;

! the Licenses shall defire and 7.5. tnose conditions and parameters in accordance

, aith 10 CFR 50.36, .hien would prevent any suen event occurring.

l t 06/01/84 18 Revision A  !

i 1

l Please evaluate the conformance with 10 CFR 50 App. A, GOC 20 and 22 of using the source Range Neutron Flux as a non-diverse reactor trip under cir-cumstances in (MCOES 3 through 5) in.wnien there is no Technical Specification on movaDie control assa.?.011es, and which instrumentation consists of only two channels. Also for circumstances in which all normally availante other Dackup t

trip functions such as pressuri:er pressure - high and low, and water level hign and " tow reactor coolant flow", are not specified to be OPERA 8LE in

! Table 3.3-1. The Licensee shall propose on the basis of this evaluation. B Items 7 & 8 Overtemperature AT and Overpower AT.

The current T.S. provides for operability of these trips in in MODES 1 & 2, and l not 3.

Occurrences using these reactor trips include events which can be initated from ,

succritical Zero Power in MCOE 2 (Reference 5. Tacle 7.2.1-4 and Reference 7, i j Taole 15.1.2 2). With the proposed T.S. In wnica no difference in Reactivity j Condition k f and Shut Down margin is required between MCOES 2 & 3. how can theLicense$j,ustifyremovalofthesetripsonentryintoMCOE3inwntenthe only difference in RCS conditions is a sarginal reduction in temcerature, from  !

the Programmed No Load T,yg.

j Item 11: Pressurizer Water Level - Hign l

Coerability considerations from M00E'2 down to and including water solid con-l ditions in the RHR MCOE are discussed under section 2.1.1 la c(if.) with a - $

proposal that exclusien of this trip for all these MODES is non-conservative in respect to 10 CFR 50, GOC 20 " Protection System Functions" both for reactivity control systems and overpressure protection systems. ,

• The necessity for this trip is increased wnen reviewed against the totality of the procosed exclusions for Reactor Trio System Instrumentation discussac in tne following section under items 2-21 (selected).

l Items 2 21 (selected):

Items 2, 5 and 6: Power Range, Interteciate Range and Source Range  ;

Neutron Flux Trips Pressurizer Pressure Low I Item 9:

Item 10: Pressurizer Pressure Hign f Pressurizer Water Level Hign Item 11:

! tem 12: Low Reactor Coolant Plow l

Item 14: Undervoltage Reactor Coolant Pumos l

Item 15: Uncerfrequency Reactor Coolant Pumos Item 21: (Proposed) Reactor Coolant Pumo Breaker Position Trio.

19 Revision A 06e01/84 l

a At this time, in MCDE 3, 4, and 5, the proposec Technical specifications for the plant do not provide any neutron flux trip for Accident Analysis require-ments, although the FSAR would require the Power-Range Neutron Flux Trip, Lew Setpoint; no insertion limits on movable control assemelies, Reactor Coolant Pump (RCP) operability requirements permitting less than four (4) ACPs in operation, a Soron concentration Control wnfen provides less snutdown margin capability than the FSAR requirements, no trip of RCPS on Loss of Flew or Undervoltage or Underfrequency or Opening of RCP breakers, and in addition it is proposed that no trip be provided for Pressurizer Pressure-Hign, Pressurizer Pressure - Lcw, and Pressurizer Water Level - High. And for these circumstances we have no well defined evaluation as to wny these recuced protections adequately I protect the plant against any of the sopropriate Condition'!!, III and IV l occurrences in these MODES except'a Large and Small Break LOCA, and Steam Line Break.

We realize the interdependence of maoy of these factors in setting a' minimum acceptacle level of Reactor Trip Protection and that relatively simple solutions are possible, cut at this time we do not have availacle an accectacle analysis and evaluation justifying the proposed T.S. position.

The Licensee shall provide an anal under acolicable Conditions !!, !!ysis and I and IVevaluation occurrencesofinthe circumstances M00E5 3 througn 5 for an appropriate set of Technical Specification requirements, to ensure conformance to Acceptable Regulatory Criteria and from this he will establish an amoropriate range of Reactor Trip System Instrumentation to Safety Related Requirements. The evaluation shall be undertaken in conjunction with our concerns for current Technical specifications unoer Sect 1on 3/4.4.1 REACTOR ~~

CCOLANT LOOPS AND COOLANT CIRCULATION of this report.

Items: 12 Low Reactor Coolant Flow Trip 14 Undervoltage - Reactor Coolant Pumps 15 Underfrecuency - Reactor Coolant Pumos 21 (Procosed) Reactor Coolant Pumo Breaker Position Trio All these Reactor Trio Functions concern potential for a loss of Reactor Coolant Flow. Th.e proposed T.5. deletas all operability requirements in l MCCES 3 througn 6. (It also deletes in MCOE 2, out this has been discussed ,

eariter under TA8L! 2.2 1 items 18.b.a and 12a and ICb]. We have discussed our related concerns and requirements for analyses and evaluations in MCCE5 3, 4 and 5 unoer Items 2-21 (selected) above.

A loss of Coolant Flow in the RCS places the plant in an Emergency Coerating Mode. Please advise therefore why such an event should not automatically trip the Reactor in MODES 3 through 5 with the 8eron Concentrations ceing consicered for the proposed Technical $cecifications. Why should we not use tne reactor trip as a device to ensure complete '.hutdown of all movaele control rods during any time that a sinimum set of RCPs in accordance with operability requirements of the T.S.. are not availante since RCPs may be required for accicent mitiga=

tion in MCOES'3 through 5 as appropriata. The Licensee shall evaluate and proDose. .

06/01/64 20 Revision A

0 0 l .

[ e l

l Item 13: Steam Generator Water Level - Low Low: e Why should not this te recuired for MODES 3, 4 and 5 (with closed locos) to  :

emersce the possibility of a return to nuclear power under these conditions. .

Further, steam Generator Cperability is also required in these Modes to remove decay heat, and Low-Low level alams are derived from the steam generator low-low instrument channels. Reference 5, Figure 7.2.1-1. The Licensee snell t i

! evalusta and propose. t Item 17: Safety Injection Input From E5F. .

See our comments on Table 2.2-1, Item 17 on a proposed revised description for *i this term to " Reactor Trip From ESFAS. ,

l The procosed 7.5. proposes that Reactor Trip on $5FAS (or 5.!) is not recuired

. to De,0PERA8LE in M00E5 3 and 4. Why is reactor trip not reouired in tnese a MODES .nen Taole 3.3-3 for E5FAS Instrumentation, and more,particularly Func-tional Unit 1, including Reactor Trip, shows operse111ty roovirements down to and including MODE 4. Further, the Itcensing basis provides that $1 including l

reactor trip, be initiated automatically and manually down to MODE a; see <

l

! Licensing Basis information in later Section 4.5, EMtRGENCY CCRE C00 LING  ;

SYSTEMS, under GENERAL, of this review. ,

j This proposed 7.5 requirement is therefore non-conservative with respect to [

l the Licensing Basis whien reautres that Reactor Trip on ESFA5 (or $1) te ,

Opertale in M00E5 1, 2, 3 and 4 The Licensee shall evaluate and propose. .

The Licensee shall evaluate the safety consequences of the fact that in the  !

I event of a Main Stream Line treek below the P-11 interlock, Reactor Trip will not be initiated by the Negative Steam Line Pressure Rate Hign signal. If  !

the break is outside containment is there is no other parameter remaining wnich  !

will cause the reactor trict if the break is inside containment will Containment '

Pressure Hign initiate reactor trip within an acceptacle time. What are the consecuences of a small to intermediate sico treak inside containment .nere. 1 i

sucn Containment Pressure High say not occur. We sopreciate that Source 4ange and Intermediate Range Nuclear Flux trips could trip the reactor uncee inese  !

circumstances, on any return to power, out their current procosed status as no-  !

Deing necessary for protection because they are not requireo in tne Safety Anal-ytes would leave only the Power Range Low Setpoint Trip, and related resulting .c power levels of 354 as a Safety Analysis Limit would be unacceptacle witnout a -

substantive analysis of the event. Please comment in terms of Reactor Trig System Instrumentation Requirements to meet these circumstances, the proposed T.5 is non canservativa in respect of Regulatory Requirements in setting inese circumstances; the Licensee snell evaluate and prosces. ,

Item: Concerning Proscribed Values For 5 RATED THERMAL P0wtR DURING STARTUP (MODE 2) AND PCwtR OPERATION (MCDE .)

I we r.ote snat ocerability reevirements for Reactor Trio System Coeration =nei  !

expressed in terms of M00E5 1 and 2 are inaccu*ata and do not recrosent the Revisici A 06/01/84 21 i

actual situation at the plant. T.S. Page 1-9, Table 1.2 defines Power Opera-tion (H00E 1) at > 5% Rated Thermal Power and Startup (MODE 2) at c 5% Rated Thermal Power. '

In actual fact.the operability positions defined in Table 3.3-1 reflect an inter-face between MODE 1 and MODE 2 determined by Permissive P-7 at a nominal 10%

Rated Power Level. Further, in this review, under Section entitled TABLE 2.2-1, REACTOR TRIP SYSTEM INSTRUMENTATION SET POINTS, item 18 c(fii) we have identified the need for Safety Analyses Limits for P-10 P-13 and in comoination for P-7, so that the outer Limits of Power level of this safety control logic can be identified for safety evaluation purposes. For example, the Safety Analyses Limit used in the FSAR for the Power Range, Neutron Flux - Low set Point is + 10%

on the set Point of 25% to give 35% as the conservative outer limit. If this same (total channel error) margin was apolicaele to both the P-10 and P-13 cnannels to give a P-7 Safety Analysis Limit of 105 + 105, i.e, 20% RATED THERMAL PO%ER, then the importance.to related safety-related tssues is -

suostantively increaser'. .

The discrepancy identified is non-conservative and important on at least 2 counts:

1) A non conservative discrepancy between the fundamental maximum T.$. Limit of 5% power level in MCDE 2 as given on T.S Page 1-9 Taole 1-2 and the i'

nominal value of 10% with a real Safety analysis Limit of 10% plus a Total Channel Error as yet unspecified.

2) The elimination of most reactor trip Functions (and many ESFAS Functions) -

at this non conservative power level without a separate comprehensive Safety Evaluation with respect to Regulatory Requirements and the existing Licens,ing Basis.

The Licensee sha11 evaluate, including our concerns encressed aoove, and propose.

i +

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3 l

I e e l

TABLE 3.3-2 REACTOR Trip INSTRUMENTATION RESPONSE TIMES ,

Item 1: Manual Reactor Trip l!

[

At this time, the Ifeensee proposes that the Response Time (RT) for manual i reactor trip is not required by safety analysis. Furthermore, he proposes that I in MODES 3 through 5, the only remaining operable trips are those using Source range neutron Flux and they also are not required by Safety Analyses.

Under TA8LE 3.3-1, items 2-21 (selected) we have already required the licensee to re-evaluate his position in respect of what neutron Flux trips he intends to propose, together with their related Tech specs to place the reactor in a safe condition in respect to Condition II, III and IV Occurrences in MODES 3

, througn 5. Until this evaluation anc proposal are accepted, tne Licensee shall nave a Safety Related Manual Trip System to assist in meeting minimum

Regulatory Recuirements in 10 CFR 50, APP A. III. Protection and Reactivity Control Systems, and the Licensee shall evaluate and propose as a priority issue. At this time, the proposed T.5 is non-conservative in respect to Regulatory Requirements for 10 CFR 50, App. A. III.

l l

! Items 5 and 6: Intermediate Range and Source Range Neutron Flux Trips. ,

l As indicated under item Taele 3.3-1, items 1-5, these items are proposed as l not being protective actions necessary for the FSAR. Analyses already requested ..  !

will provies a base for determining whether those trips are necessary to pro-l tect the plant in MODES 3 through 5. If so, please provide the necessary techn-l ical specifications for these response time in confomance with 10 CFR 30.46.

l If these values are not provided, all related return to reactivity. events shall l

be evaluated by the Licensee with current FSAR reoutrements for the Safety '

Analyses Limit of the power range, neutron flux, low setpoint trip wnten will l be required to be OPERASLE.

l The current preposals for these trips is non conservative with respect to l other proposals in the f.$; the Licensee shall evaluate and procose.

Item 8: Overpower AT. ,,

No response time is provided by the Licensee who proposes that a T.$. on this is Not ApD11ca01e.

Please comment on the fact that this reactor trip is proposed in Referenes 5 Taele 7.2.1 3 (3 of 5) as applying to five (5) separate condition II tnrough IV licensing basis occurrences. Also that Reference 5, Page 7.2 14 Rev.42.

I item 1 d) specifies a maximum of 6.0 .econds (including a trtnsport time of '

! 2 secs) and which is confimed by Reference 7. Taole 15.1.31 (alongsiae l Overpower AT).

l The proposed T.$ is non conservative with respect to the Licensirg Basis, the l Licensee shall evaluate and propose, r l

Item 9: Pressurizer Pressure Low

] \

L 06/01/84 23 Revision A l l

l 1

! I f

4{ Item 10:Pressurizer Pressure - High

- %g The TS specifies a Response Time of 12.0 secs. Reference 7, Table 15.1.3-1 l provides a time delay of 2.0 secs for these events wnich conflicts with a value of 1.0 secs in Reference 5, page 7.2-14, rev. 42, item 1(e). The Licensee shall clarify.

Item 11: Pressurizer Water Level - High No response time is provided because it it considered Not Applicable (NA).

The trip is shown as having a protective function for two Condition II occurrences in Reference 5, Table 7.2.14 (4 of 5) and a potential protective function in a Condition IV occurrence in Reference 7 page 15.4-13, item 16 c.

Accitional item 11. protective functions are, discussed earlier unde.r Tacle 2.3-1, Reference sponse time5,atpage 1 sec.7.2-14, Revision 42, Item 1 f provides a reactor trip re-t Reference our earlier review under Table 2.2-1, item 18.c.(fi).

In view of the above information, the proposed T.S. is non-conservative with respect to the Licensing Basis. The Licensee shall evaluate and propose.

Items 8 & 11 General -

Although the above two items are not apparently the primary reactor . trips used as the basis for calculating protection in the Accident Analyses in reference 7, those Analyses represent a limited nuncer of events wnich are proposed as

' expected" to bound all possinle events at the plant in terms of severity.

There is no guarantee that the large number of other possiD1e events will never use these two protection items to primary aavantage.

Item 16. Turtine Trip A response time for Reactor Trip on Turnine Trip is not orovided in the Tecnnical Specifications. Reference 7, Tacle 15.1.3-1 advises tnat tne re-sponse used.

analysis time for such a trip is 1.0 sec. but that it is not applicable to the ,

Reference 7. Section 15.2.10.3, concerning Excessive Heat Removal Oue To Feedwater System Malfunctions. Under the title of "Results" on page 15.2 30, the second paragraph describes how for this particular event at full power "A turoine trip and reactor trip are actuated when the steam generator 1 .e1 reaches the high high level set point."

Also, for the Occurrence of "!nadvertent coeration of the ECO3 Quring Power Operation under reference 7, Section 15.2.14.3, page 15.2 40, revision 43, under Conclusions low pressure reactorstates trip isthat: "!f the reactor does not trip immediately, the actuated. This trips the turoine and prevents excess cooldown thereoy expediting recovery from the incident.

06/01/64 24 Revision A

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Under these circumstances therefore, Reactor Trip on Turbine Trip is necessary to automatically terminate the event. The Licensee should review the response time used in the above calculation and provide an evaluation of its decision is respect of placing it in the T.S. under the requirements of 10CFR50.*6 Item 17, [ Reactor Trip on] Safety Injection Input from ESF This description is a misnomer and should be replaced by the description proposed under Table 2.21, Item 17 of this document.

4C [ The proposed T.S. states' that the response time requirement is NA (Not Applic-acle). This is incorrect as a separate Reactor Trip is an essential part of all ESFAs functions during wnich safety injection is initiated. The required infomation is in f act supplied in T.S. Page 3/4 3-30 Table 3.3-5, under the alreacy revised headings proposed above, reference items.11, Ob, 3b. ab.

~

This table, under response time, should replace the description as. recommended acove and alongside each, reference the entry in T.S. Taole 3.3-5.

The response given in the Technical Specifications (except for Manual actuation of SI) are quoted as < 2 secs. No docketed information is available on what values were used in accident analysis, and particularly for MSLS, SBLOCA and LOCA events. The licensee should provide this information and confim its conservatism against the T.S. value, 49. reference 5. Tacle 7.2.1-4 (5 of 5) and related note e. on page entitled " Notes for T401e 7.2.1-4" confirms' that Pressurized Low Pressure - Low Livel is the first out trip of Safety Injection for the event of " Accidental Dec.aessuritation of the Main Steam System." The licensee sna11 explain this terminology - wnether we have Reactor Trip on Pres-surizer Pressure - Low which is availacle 'at the maximum power output at which this particular event is esa.luated, or I?ressurizer Pressure - Low (Safety Injection) and provide the associated response time to validate proposed T.S.

vaiues.

Item 21, Procosed (Reactor Coolant Pump Breaker Position Trip)

As discussed earlier, under table 2.21, Item 14, this trip is proviced as an adjunct to Undervoltage - Reactor Ooolant Pt.mp Trip. The Licensee snail evaluate and cropose.

06/01/84 25 Red ston A

-.. - .: u - . .=- . . a .- _ .:. . . .: -- .-- . .. . : =.... .:- . . . - -. . .

TABLE 3.3-3 ENGINEERED SAFETY FEATURES ACTUATION SYSTEM (ESFAS) INSTRUMENTATION Item 1: Safety Injection, Reactor Trip, Feedwater Isolation, Component Cooling Water, Start Diesel Generators, and Nuclear Service Water. )

This description of Item 1 lists the various functions initiated by given signals (which are generally those initiating SI).

However, Reference 5, Figure 7.2.1-1 (8 of 16) revision 34 and Figure 7.2.1-1 (13 of 16) revision 34, shows that the term "Feedwater Isolation" used in this Item 1 is actually comprised of four (4) separate Logic Functions, namely  !

"Turnine Trip", " Trip of Feedwater Rumps", "Close All Feedwater Isolation i Valves" and "Close'the Feedwater Main and Bypass Modulating Valves.

The term Feedwater Isolation is therefore an inaccurate term to use. It should De removed frem.this cescriptor and replaced by the four separate functions, as each of them can be initiated separately and or together dependent ucen the initiating Logic.

Further we also note that this functional unit is also that initiated by Steam Generator Water Level High-High (P14) reference 5, figure 7.2.1-1 (13 of 16) revision 34. and figure 7 of 16; revision 41.

Further, the function to be initiated by Steam Generator Wate'r Le'v el - High

Hign is function 5 of the.same Table which is again incompletely described and should be changed (see item 5 later) to clearly identify these same 4 elements. ..

Under these circumstances, the current description for Item 1 should delete the term "Feeowater Isolation" and Item 5 (see later) should be expanded to include an additional Functional Unit identified as Safety Injection.

Additionally, the Function " Annulus Ventilation" needs to be added to the descriptor (reference 5. figure 7.2.1-1 (8 of 16) revision 34).

Also, the function unit cescription " Nuclear Service Water" should include

[ isolation and startup] of Nuclear Service Water.

Item la): Manual Initiation This should read as: Manual Safety Injection Actuation. [There is not a ,,

separate Manual Actuation for each of the functional units listed.]

Item ic: Containment Pressure - High/ Applicable MCDES 1, 2, 3.

The Current T.S. does not provide for initiation of SI en Containment Pressure - High, in MODE 4 This is contrary to reference 8, pages Q212-47e, item 24, Q212-61b item 29, Q 212-61d, item 212.91 (15.4) wherein small and large breaks in the Steam Line and Reactor Coolant System are discussed down to and including MODE 4. Discus-sing NON-LOCA Accidents (in MODES 3, 4) below the P-11 (1900 psig) block of SI on Pressurizer Pressure - Low (SI) and Steam Line Pressure - Low, provision is, made that if a MSLS occurs inside containment [so that MSIV Isolation on 06/01/84 ._ 26 Revision A n ,-~,--,-,--,.-e.p,-%-w_-e , <eg gy 3-y ep-,- . .e, <_.-%-gs .G ev- __ _ - - ,-

t Negative Steam Line Pressure Rate - High does not contain the event for the Faulted SG) then Safety injection will be activated by Containment Pressure-High.

Note: Automatic logic for realignment to SI is already provided in the T.S. in MCDES 3 and 4. This MODE 4 Cperability requirement for Containment Pressure-High would also facilitate re-alignment of equipment from RHR to ECCS alignment in the event of a large break LOCA under these circumstances as described in reference 8, page Q212-47a, item II.C.

The Licensee shall evaluate why his proposed T.S. is an acceptable change frem the existing Licensing Basis, or incluce the operability requirement in his T.S.

The proposed T.S. position is non-conservative.

Item la: Pressuri:er Pressure-Low This is the same title as used for Reector Trip on Pressurizar Pressu're-Low.

This particular/ESFAS actuation is set at a lower pressure and should be described as: Pressuri:er Pressure-Low [ Safety Injection].

Item le:

The proposed T.S. for SI on Steam Line Pressure - Low is qualified in MODE 3 by a 3## which is identified on T.S. Page 3/4 3-23 as a situation in which the function may be blocked below P-12 (Low-Low T,yg Interlock) setpoint.

Reference 5, Table 7.3.1-3 (1 of 2) and (2 of 2) item P-1, shows the appropriate -

interlock for this purpose is P-11. Item P-12 of the same Table makes no provision for this proposed T.S. position.

However, reference 5 figure (6 of 16) does not use the same manual block (at P-11) for Pressuri:er Pressure - Low (SI) as for Steam Line Pressure - Low (SI) (and imolementation of Negative Steam Line Pressure Rata) on reference 5.

Figure (7 of 16). The Licensee is required to confirm that no parameter other than the value of Pressurizer Pressure (at P-11) is used to concition the manual blocxs relating to the steam line; if other parameters are used, the Licensee shall evaluate and propose. The Licensee shall also acvise of otner parameters which may be used to condition the manual block of Pressuri:er Pressure - Low (SI). ,

If the Taole 7.3.1-3 (1 of 2) and (2 of 2) is correct, then condition MODE 3## shoula be changed to condition MODE 3# which becomes tne correct cescription.

Item 2c: Containment Pressure-High-High.

Operability is not required in M0ud 4. This should be required to be consistent with the evaluation under Item 3.b.3. below.

Item 3.b3): Cor.tainment Phase B Isolation on Containment Pressure - Hign Hign Operacility of this isolation is not proviced in MCDE 4. The Licensee should aavise wny this is not necessary for safety when the previous item No.l.e.

06/01/84 27 Revision A

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showed reference in the Licensing Basis of protection against Steam L'ine Break inside containment and Large Break LOCA in this mode. It should be noted that T.S. Item 3.4.6.1 requires containment integrity in MODES I.through 4.

Further Operacility of Auto-Actuation Logic is required through .MCCE 4 [Contain-ment Pressure-Hign only effects Containment Isolation A and not Containment Isolation B which is necessary to establish Containment Integrity].

The proposed T.S. is non-conservative. The Licensee shall evaluate and propose.

Item 3c: Purge and Exhaust Isolation An additional Item: 3c.4 Containment Radioactivity, is proposed to effect Purge and Exhaust Isolation as this is part of ESFAS Logic in reference 5, figure 7.2.1-1 (8 of 16), revision 34 The Li' censing Basis for this requirement lies inside the analysis of consequences deriving from accidental events wniist the Purge and Exhaust Isolation Valves are open. [Refce CSB]

The proposed T.S. is non-conservative with respect to the Liceicdeg Basis; the Licensee shall evaluate and propose.

Item 4, Steam Line Isolation 4b: Automatic Actuation Logic and Actuation Relays The p'roposed T.S. does not require Operability of Steam Line Isolation Auto -

Actuation Logic in MODE 4. However, this will be required if the Operability requirements of Steam Line Isolation on Negative Steam Line Pressure Rate -

High, already specified in item 4d for MODE 4, are to be met. The proposed .T.S.

is non-conservative with respect to the Licensing Basis; the Licensee shall evaluate and propose.

Item 4a: Manual Initiation (of steam line isolation)

1) System

2) Individual Operability, requirements for manual initiation of Steam Line Isolation are not required by the current T.S. in MODE 4. This however will be necessary to -,

allow the operator to manually isolate small breaks which do not activate the Negative Steam Line Pressure Rate - High signal or the Containment Pressure-High High signal.

The proposed T.S. is non-conservative with respect to the Licensing Basis; the Licensee shall evaluate and propose.

Item 4d: Negative Steam Line Pressure Rate - High Operability recuirements are given as MODE 3 and 4. MODE 3 should be con-ditioned as MODE 3# indicating it is only available below P-11 Interlock.

The Licensee shall evaluate and propose.

06/01/84 28 Revision A O

._ _ . a . . c_. . . . . . . - . ~ . . . _ _ . . . 1 Item 5: Turoine Trip and Feedwater Isolation Reference earlier Item 1 in which this title for Item 5 snould be more accurataly described as " Turbine Trip, Trip of Feeewater pumps, Close Feedwater Isolation Valves, Close Feedwatar Main and Sypass Modulating Valves. The Licensee shall clarify, evaluate and propose. Lack of accuracy can be non-conservative with respect to the Licensing Basis.

Item Sa: Automatic Actuation Logic and Actuation Relay [to effect Turbine Trip, Feedwater Pump Trip, Closure of Feedwater Isolation Valves and Closure of Feedwater Modulating Valves]/ APPLICABLE MCDES 1 & 2 The Applicable Modes of this Auto Actuation Logic need to be extended down to MODES 3 and 4 to be available to respond to the Safety Injection signals whien are expected frem the Licensing Basis (reference later Section 3/4.5,

Emergency Core Cooling Systems, under GENERAL). The preposed T S. is non .

• conservative with respect to the current Licensing Basis and the Licensee shall evaluate and propose.

Item Sb: Steam Generator Water Level - High High [to effect Turbine Trip, Feedwater Pump Trip, Closure of Feedwater Isolation Valves and Closure of Feedwater Modulating Valves]/ APPLICABLE MCOES 1 & 2.

I The Licensee should evaluate the need to extend the operability requirements of this functional unit from current MCDES 1 and 2 down to and including MODE 4 The determining factor may be the availablity of Main Feedwater Pumps during -

these MODES. Plant Operating Procedures which permit Main Feedwater Pumps to be available can cause An Excessive Heat Removal Oue To Feedwater System Mal-function and/or Steam Generator overfill unless Safety Related isolation at the Main Feedwater [ containment] isolation valves is incorporatad into the T.S.

The Logic of reference 5, figure 7.2.1-1, (13 of 16), revision 34, involving signal inputs: Steam Generator Hi-Hi P-14, Safety Injection, Reactor Trio 24, and Low T,y would need to be carefully reviewed, especially since there is currently little or no Safety Related Reactor Trip Protection in MCOES 3 through 4 so that reactor trip P4 may not be available in conjunction with L:w T

avg (during cc id wn) to effect Feedwater Isolation, and Closure of Moculating Valves, as an inouilt protection against such circumstances.

The preocsed T.S. does represent a non-conservative position in respect to the Licensing Basis, as there is no prerequisite that Main Feecwater is isolated a:

.the Containment Isolation Valves as an LCO, during MCDES 3 and 4 The Licensee shall evaluate and propose.

Item Sc (Proposed): Safety Injectic. [to effect TurDine Trip, Feecwater Pume Trip, Closure of Feedwater Isolation Valves and Closure of Feecwater Modulating Valves]/ Applicable MCDES PROPOSED AS 1, 2, 3 and 4 This trip is relocated from Functional Unit 1 to Functional Unit 5 in accordance with our earlier reviews under Item IC and Item S.

06/01/94 29 Revision A n- . . _

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OPERASILITY is required in all Modes 1, 2, 3, 4, because SI protection has been found necessary within the Licensing Basis. The protection was already intended in the procosed T.S. this action represents a more accurate description of the Functional Unit and an improved placement in the T.S. The Licensee shall evaluate and propose.

Item 7; Auxiliary Feedwater (AFW):

General: Operability Requirements:

Requirements for ESFAS operability in AFW are generally ifmited to MODES 1, 2 and 3. However, provision is made in the FSAR for operation in MODE 4, and to be available in MODE 5. .

For MODE 5, Reference 8 page Q 212-56 rev. 25 where RCS cooling is required to be availabit in the event of failure of on,e of the isolation ~ valves in the'line leading from the RCS hot leg to the suction of the RHR $ causing flow blockage. Available Operability during MODE 5 is necessitated to facilitate conversion to effectively MODE 4 operation, as described in

' . reference 8, page Q 212-56, rev. 25, since "only a few minutes" is pro-posed as necessary "to open the steam dumps and to start up the auxiliary feedwater system." It is proposed by NRC, that such a rapid startup of the AFW system can only be achieved by having available the Automatic Actuation Logic and Actuation Relays, and all related ESF equipment so that the automatic logic can be initiated manually. The licensee shall evaluate and propose. The proposed T.S. items 7a through 7g are gener-ally non-conservative with respect to the Licensing Basis in this matter. ~-

! The Licensee shall evaluate and propose on each of these items including l-consideration of our related reviews.

Operability in MODE 4 is required by the FSAR to generally counter the consequences of appropriate condition II, III and IV occurrences including Steam Line and Feedwater Line Breaks, which are analyzed assuming automatic ini tiation.. Reference also procosed T.S. pages 3/4 4-3 for requirements for operable RCS systems in MODE 4 The proposed T.S. ftems 7a througn 7g are generally non-conservative with respect to the Licensing Basis in this matter. The Licensee shall evaluate anc propose on eacn of taese items, including consideration of our related review.

Item 7.a: AFW/ manual initiation Item b: AFW/ Auto Actuation Logic and Actuation Relays:

Operability is currently not required in MODES 4 and 5. Operability should be provided for both modes to meet the licensing requirements, i.e., manual initiation of Automatic Actuation Logic and Actuation Relays: reference General above.

Item 7.c.1: Start Motor Driven Pumps:

Should be operable in both MODES 4 and 5 and especially to counter non-availacility of Turbine Driven Fumps early into MODE 4 during the coolcown.

05/01/84 30 Revision A e

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Item 7.c.2): Start Turbine Driven Pumps: ,

Should be operable in 4. Although not capable of operating at lower tem-paratures of MODE 4, and MODE 5, it should nevertheless be available for use to counter consequences described in " General" above, including a station blackout.

Item 7.d): Auxiliary Feedwater Suction Pressure Low:

This proposed T.S description of a functional unit is invalid. The Functional Unit to be provided is:

d) Automatic Re-alignment of Suction Supply [This is the functional unit],on Low Auxiliary Feedwater Suction ~ Pressure [This is the parameter caus-ing the change] -

Operability requirements should identify how many AFW pumps are required to be " tripped" deficient in suction, to effect re-alignment.

The licensee should identify those instrument / control channels, and partic-ular engineering alignments, which result in a re-alignment of reduncant AFW supplies to the only safety-related supply available, from the Nuclear Service Water Pond, and define related operability and surveillance require-ments. The mixed nonsafety and safety-related supplies on the McGuire y units make it necessary to separately define and T.S. those safety-related elements, under 10 CFR 30.46: see reference 14, page 10-2.

Applicable Mod'es in the current T.S. is limited to 1, 2 and 3. The licensee shall evaluate why this should not be extended to MODES 4 ana 5 to meet the FSAR requirements described in " General" above.

Item 7.e: ' Start Motor-0 riven Pumos (by Safety Injection)

Apolicable Moces have not been identified. NRC proposes MCCES 1, 2. 3 anc 4 and 5 to meet tne requirements of Item 7: General, discussec earlier.

Item 7.e: Start Turoine-Oriven Pumps (by SI) l This functional unit proposes that the Turbine Driven AFW pumps are startaa by the SI signal. This conflicts with reference 5, Fig. 7.2.1-1 (15 of i 16) I&C system Logic Diagram where the initiation of~the turbine driven pumos on SI is not shown. Also, in a like manner, with related sec-tion 7.4.1.1.1.1. and reference 22, section 10.4.7.2.2.6. Also see refer-ence 14 Section II.E.1.2 page 22-41. It is now noted that the recent T.S. has been corrected is show that the Turbine Driven AFW pump does not start on Safety Injection.] The Licensee shall clarify.

1 06/01/84 31 Revision A

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s Item 7.f; Station Blackout - Start Motor Driven and Turbine Driven Pumps:

Provision for operability is only in applicable MODES 1, 2 and 3. Con-sistent with previous considerations, operaoility should be required in MODE 4, with provision for immediate operability from MODE 5.

Item 7.g: Trip of Main Feecwater Pumps (MFWP) - Starts Motor Driven Pumps The T.S. proposed only 1 channel per pump to trip. [This is different to the FSAR, reference 22, page 10.4-14, rev. 7, item 30 which specifies that loss of g main feecwater pumps is required. The licensee should evaluate and pr; pose. -

So- Acolicable modes: The current T.S. proposes Modes 1 and 2#. Condition 2#

is an invalid MODE since # identifies the P-11 interlock which can be manually effected only at approx. 1900 psig and which can only occur in MODE 3, i.e., the condition shou'd b'e 3#.

' The licensee.should explain and propose.

Please advise why this limitation at MODE 2 [or 3]# is proposed and how it may relate to plant operating procedures in MODES 3 and 4 and whether this block is in conformance with regulatory requirements.

Item 8: Automatic Switchover to Recirculation on RWST Level: . b This is limited in App 1icability to MODES 1, 2, 3 by the proposed T.S.

l Since a LOCA in MODE 4 is part of the Licensing Basis, see later Sec-tion 3/4.5 ECCS under GENERAL, the licensee should evaluate the reasons

{ for, and the consequences of, not proposing this OPERA 3LE IN MODE 4, and not being available in MODE 5, to counter the consequences of potential LOCAs and loss of RHR cooling in these MODES. The proposed T.S. is non-conservative with rescect to the Licensing Basis; the Licensee shall evaluate.and propose. J Item 9: Loss of Power: Emergency 3us Undervoltage - Grid Degrace voltage:

l l Item 9: General The Licensing Basis FSAR, reference 7, Section 15.2.9 under LOSS OF 0F: SITE POWER TO THE STATION AUXILIARIES describes a set of Reactor Protection System and Engineered Safeguards Features Actuation responses for the plant to ensure its safety. Why is this particular set of ESFAS Func-tional Units and related Response Times not provided under Taole 3.3-3.

Absence of this information makes the proposed T.S. non-conservative.

I The Licenses shall evaluate and propose.

What does this functional unit do. Please explain, and how many busses to l be tripped for the action to be defined. If it is meant to initiate AFW:

wnat pumps etc., and if so operability requirements should be extended to MODE 5. Lack of any clarity makes this proposed T.S. non-conservative.

The Licensee shall clarify, evaluate and propose.f 06/01/8.r' 32 Revision A

.. 2. = ==: ...

.. . . ==w : -

2- - - - -

Item 10.pi.: Pressuri:er Pressure P-11:

Applicaole MCDES are 1, 2, 3.

Explain the consequences of this non-operability in MCDE 4 on availability of dependent protective actions, e.g., main steam ifne isolation, which is considered under Item 4.b above. If main steam isolation is negated, it should be restored to conform to Regulatory Protection Requirement. The 1

Licensee shall evaluate and propose. ,

! Concerning P-11 Interlock and AFW Pumes.

The basis provided on proposed T.S. Page B 3/4 3-2 states that:

"P-11 (i.e. , on system pressure increasing to P-11 valve) ---- Defeats the manual block of the motor driven AFW pumps on trip.of, the main feec-water pumps and Low-Low Steam Gene'rator level."-

The following information provides the current Licensing Basis on the particular proposed interlock P-11 in respect of AFW Pumps:

The Table 3.3-3, Item 7.c.1, in reference 5. for start of motor driven AFW pumps, does not provide for the above condition.

The P-11 interlock and its provision for automatic defeat (above P-11 setpoint]

do not appear in reference 5, Tacle 7.3.1-3. Rev-35 Interlocks for ESAS and -

Figure 7.2.1-1 (15 of 16), revision 34 I&C Logic Otagram.

Reference 5, Section 7.4.1.1.6 describes this action under " Bypasses anc Interlocks" and that whenever it is present, an alarm exists in the Control Room. This allows the operator to stop AFW pumps curing snutcowns.

Sucplement No. 5, reference 15, page 22-22 evaluates the use of the P-11 in er-lock as described in the above Basis and concludes that the situation is acceptable. However, the basis for the SER Suco 5 conclusion was that a :ossi-ble steam line ructure or feedwater line break were not likely to occur in the proposeo MCDES when the P-11 is in effect. This is a mistake, all the earlier

' work of this review has disclosed that the premise of these events being net ,,

likely to occur has been rejected for these MODES 3 to 5, and detailed attan-tion has been given to their possible occurrence together with the possibility of Auto Initiation and the consequences of automatic protective action. Where the P-11 lockout nas been present on other protective actions, the consecuencas nave been fully evaluated. There has r.ever been a.related evaluation on the absence of auto-initiation of motor-driven AFWS as now proposed.

If the Licensee wishes to pursue *lis he should evaluate all the events considered in the FSAR below the P-11 seccoint with manual initiation of MD AFW and making due allowance for all the relative reduced and changed protections available and the time frames which must allow for all other actions, e.g.,

isolation of a ruptured SG is expected to take 30 mins, see reference 7 section 15.4.2.2.2 page 15 4-13a, Revision 38. Further, the cetailec review of this T.S. has been b'ased on this availability.

Revision A 06/01/84 33 1

py , , , _ _ _ -"T; - _ _ . . _ _ , _ __ -

~

• ~_

s We. note that in his submittals concerning this matter, dated March 9, 1981 concerning TMI items, the Licensee states that "the turbine driven auxiliary feedwater pumps do not have a bypass feature." Yet we also note on his T.S.

page 3/4 7-4 that the Turbine Driven pump is not required to be operable when steam generator pressures are less than 900 psig; this would require only approx. 20 mins. into stancby cooldown to achieve. The result is that there would be absolutely no automatic supply of feedwater for any event beyond approx. 20. min into cooldown.

At this time, the current Accident Analyses in the Licensing Basis FSAR support the necessity for not using the current bypass for the Motor-Driven Pumps.

Il The Licensee shall advise wha.t safety-related reasons require that he mus bypass automatic startup of the motor-driven auxiliary feedwater pumps on l:

top ofshutdown.

plan,t both main feed pumos, and on SG Low Low-Level in the final stages of Also, what prevents him from installing automatic restoration

on receipt of the related protection-signal.

Item 10.b; Interlock; Low-Low Tavg P-12:

Applicable MODES are 1, 2, 3.

Reference Item Table 3.3-4, Item 10b, of this document.

Since Interlock P-12 effectively provides and limits steam dump capability, including accidental blowdown, by constraining it to 3 cool down dumps to the condenser; why remove this interlock in MODE 4 and MCDE 5 and remove its potential availability for related Licensing Basis requirements. The proposed T.S. is non-conservative with respect to the Licensing Basis; the i

Licensee shall evaluate and propose.

l Item 10.c; Interlock; Reactor Trip P-4:

The eight' separate functions affected by this interlock are described in reference 5, Table 7.3.1-3 (1 of 2). Please evaluate how the ausence of this will affect the various functions to be performed and how they will imcact the FSAR requirements for plant protection in MODES 4 and 5. This should be for both the " Reactor tripped" and " Reactor not trippec" conci-tions consicering that the reactor can be in both situations during these Modes. Licensees evaluation to items 5a, b and e acove shoula be also

! consicered in this evaluation.

i l

!- The proposed T.S. is non-conservative with respect to the current Licensing Basis. The Licensee shall evaluate and propose.

Item 10.d); Interlock; Steam Generator Level-High High, P-14:

Operability is not required by the T.S. in MODES 4 and 5. The need for this interlock in these Modes will be established by the Licensee in his response to items Sa, b and c above. The licensee shall provide his svaluation and propose. Until Safety Related Isolation of Main Feecwater 06/01/84 34 Revision A

Containment Isolation Valves is included in the T.S., this proposed T.S. "

must be considered non-conservative with respect to Regulatory Requirements.

Item 11 proposed:

There is a need to add a new Functional Unit not addressed in the current 3 T.S., but which is a part of ESFAS.

This is:

"Close All Feedwater Isolation Valves" and "Close the Feedwater Main .

and Bypass Modulating Valves" I See reference 5, Figure 7.2.1-1 (13 of 15) revision 34 for the related unique control logic. ,

This Function is initiated by:

r lla. Reactor Trip P-4, and Low Tavg.

lib. Reactor Trip P-4, and Steam Generator Level - High Hign P-14 11c. Steam Generator Level - High High P-14 (see 5 above)' .

lid. Safety Injection (See 5 above). -

Operability for 11a would be in accordance with lac (above) and later evaluation under Table 3.3-4 Item lia (Proposed). Operacility for 115 -

would be in accordance with the evaluations in 10c and d above.

Operability for 11c and 11d would be by reference to items 5, Sabc.

TABLE 3.3-3: TABLE NOTATION The uneartainty of the notation under ## is discussed in Item le earlier.

Please amend as required in accordance with the related resolution.

.M 35 Revision A 05/01/84 e

~

TABLE 3.3-4: ENGINEERED SAFETY FEATURES ACTUATION SYSTFM (ESFAS)

INSTRUMENTATION TRIP SET POINTS General: These have been checked against the information in reference 13, tacie 3-4 and related NOTES FOR TABLE 3-4 on page 3-13 and which is de-scribed as being applicable to McGuire Unit 1, 50-369. At this time, the assumption is made that this information also applies te McGuire Unit 2,.

Docket No. 50-370. The licensee will docket this fact or otherwise docket the alternate information.

Item No.1: -

The description for this Functional Unit should be clarified and modified in accordance with our remarks under TABLE 3.3-3; Item 1.

av . .

ItemNo.gy: .

The description for this Functional Unit should more accurately read as " Manual Safety Injection Actuation." See reference 5, Figure 7.2.1-1 (8,of 16),

Revision 34 .

Item Id -

s Modify the description in accordance with our earlier comment under Table 3.3-3 Id to: Pressurizer Pressure - Low (Safety Injection)

Item 3c.4 (Proposed): -

, Reference 5, Figure 7.2.1-1 (8 of 16) revision 34 shows that " Containment Radioactivity" initiates containment ventilation (Purge and Exhaust) isolation.

Please explain why it is not included as, e.g., a proposed Item 4). The pro-posed T.S. is non-conservative with respect to the Licensing Basis. The Licensee sna11 evaluate and propose.

h4 -

]

I Itam 4d: Negative Steam Line Pressure Rate - Hign [For isolation of the MSIVs below P-11 Block]

The trip set point is currently specified at -100 psi /sec. Westinghouse Set Point Methodology for Unit 1, reference 18, shows this value tc be

"-110 psi"; an additional descriptor is also necessary reading: "witn a time constant of 50 secs". The current " Allowable Value" in the T.S. is

-120 psi /sec, the same reference 18 Table 3-4 shows this value to be -100 psi; this should again have the additional descriptor reading: "with a time constant of 50 secs".

To discuss negative values and related conservatisms, it is clear to delete the - in -100 as the description reads : " Negative Steam Line Pressure Rate - High so that T.S. values should read as 100 osi and 110 osi. This is also internally consistent with the descriptor in Tacle 2.2-1. Item 4, namely: Power Range, Neutron Flux High Negative Rate, 5%

of R.T.P with a time constant of 2 seconds.

06/01/84 36 Revision A-

Please discuss the logic of the values in reference 18. A Trip Set Point '

of a negative rate of 110 psi with an allowable value of 100 psi (both with a time constant of 50 psi) would provide that an earlier isolation of the MSIVs is less conservative, and this is not so for the MSL3 event.

The excectations are that nagative rate for the allowable value would be higher than for the Set Point. Please clarify.

Further, the same reference 18 Table 3-4, column 12, states under i notation (5) that this value is not used in the safety analyses. Since this ESFAS signal provides Main Steam Valve Isolation on Main Steam Line Break below the P-11 block point (instead of by Steam Line Pressure - Low) .

please describe how the plant is othemise protected through the procosed T.S. Othemise, please provide analyses which show that the plant is pro-tacted by this proposed setting under' proposed T.S. requirements. This item is related to our other concerns on Technical Specifications on Bora-tion Control under. earlier Section 3/4.l.1 Boration Control. The proposi-tion that this value is not used in Safety Aanlysis is non-conserva'tive.

The Licensee shall evaluate and propose.

Item 5: The description of this Functional Unit should be revised and clarified to our recommendations under Table 3.3-3, Item 5.

Item Sc: Proposed new item as " Safety Injection" This should be included in accordance with the evaluation under Tacle 3.3-3, Item Sc) -

Item 6a & b. Containment Pressure Control System The lice ~nsee should provide the basis for these Set Points and Allowable Values.

Item 7(c): ' Steam Generator Water Level - Low-Low The licensee should rescend to our concern under Tacle 2.2-1, item 13.

Item 7(d): Auxiliary Feedwater Suction Pressure Low The description should be revised as proposed uncer our earlier Taole 3.3-3 item 7d. Provide the basis for the values given.

F Items 7c(1) and (2): Conc,: ofng start of Motor Driven and Turbine Driven Pumos This technical specificatinn provides that the motor-driven AFW Pumps start on low-low in one SG whereas the turcine driven pumps require low-low in two SGs. This appears to be in conflict with the accident evalua-icn in the Licensing Basis FSAR as elaborated below. [This however is not conflict with the Instrumentation & Control Logic of the FSAR.]

06/01/84 37 Revision A

(a b (W-? * -})

f (

Item 7c:

Reference (7) related Section 15.4.2.2.2 concerning Main Feed Line Rupture (MFLR) under the title of Major Assumption 10.

"The auxiliary feecwater system is actuated by the low-low Steam Generator Water Level Signal. The auxiliary feedwater system is assumec to supply a total of 450 gpm to three intact steam generators.

k -

Reference 5, Section 10.4.7.2.2 states that " Travel stops are set on the steam generator flow control valves such that the turbine driven pump can supply 450 gpa to three intact steam generators while feeding one . faulted generator and both motor driven pumps together can sucply 450 gpm to three intact steam generators while feeding one faulted generator. The throttle positions allow all three pumps to supply a total flow of 1400 gpm to 4 intact steam  ; generators."

Reference 7 related Section 15.4.2.2.2, page 15.4-13a (Revision 38),

i i

states: "The single active failure assumed in the analysis is the turoine driven auxiliary feedwater pump. The motor driven pumo that is headored to the steam generator with the ruptured main feedline supplies 110 gpm to the intact steam generator. The motor driven pump that is headered to two intact steam generators supplies 170 cpm to each. This yields a total flow of 450 gpri to the intact steam generators one minute after reactor trip. At 30 minutes following the rupture, the operator is assumed to isolate the auxiliary feedline to the ruptured steam generator which results in an increase in injected flow of 80 gpm." '

'The sequence of events in the accident evaluation in Reference (7),

Table 15.a-1 shows that after the accident is initiated at a programmed value of SG 1evel, the low-low SG 1evel in the ruptured SG is reached

'l 20 secs. later, and auxiliary feedwater [at 450 gpm] is delivered to the intact steam generators in 61 sec.

It appears, based on the above information, that on SG low-low in the ruptured SG, both the motor driven and the turoine driven pumos are i initiated (with the single failure being in the turoine driven pumps).

This is not in accord with the T.S. If it is assumed that low-low level j in the other SGs is also reached at the same time by bubble collapse, plesso justify. We note that the Reactor & Turbine Control. System is cesignea so that under normal operation, collapse of SG 1evel on Turoine I

Trip will not cause a reactor trip; also at this time, main steam from intact SGs is being lost to the faulted SG so that whereas inventory is lost, a full collapse need not occur.

(#1 The proposed T.S.s 7cf and 7.c(2) appear to be non-conservac1ve in respect )

' of Accident Analysis used in the Licensing Bases. The licensee shall j clarify, evaluate and propose $this snould be in conjunction witn our otner concerns on tnis event noted later in Sections of this review.

06/01/84 38 Revision A i

Item 8: Automatic Switchover to Recirculation The Licensee shall provide the basis for the set point values of the RWST i levels specified. What are the allowable values for [ drift and] total channel errors and _the related Safety Analysis Limit.

hC Item 9: Loss of Power I

Confirm the bases for the set points and allowable values specified.

k s Item: General The Licensing Basis FSAR, reference 7, Section 15.2.9 under LOSS OF OFFSITE POWER TO THE STATICN AUXILIARIES describes a set of Reactor ~

Protection System and Engineered Safeguards Features Actuation Responses for the Plant, to ensure,its safety. Why is.this particular set of ESFA:s Functional Units and related- Instrumentation Set Points not proviced.in

' ~

this item under Table 3.3-4?

Absence of this information makes the proposed T.S. non-conservative.

The Licensee shall evaluate and propose.

Item 10a: ESFAS Interlock Pressurizer Pressure, P-11.

Actuation of this interlock substantively reduces ECCS protection against Conditions II, III, and IV Accidental Occurrences.

The FSAR has analyzed the consequences of this reduced level of protection for a limited number of these occurrences and this has been ba:ed on a system pressure of 1900 psig; Reference 8, page Q212-47, item 212-75 1A.

Why then is a trip set point of <1955 psig used. This set point value should be below 1900 psig with appropriate allowances for drift and enannel errors to the limiting value used in the Safety Analysis of 1900 psig. The current specification is non-conservative with res ect to the Licensing Basis FSAR & therefore not in accordance with 10 CFR 50.35. The licensee snall provide a safety evaluation for the difference, for a:Oroval, or restore the set point to be a valid T.S. value.

Item 10b: ESFAS Interlock T,yg-P 32 ,

The basis for this interlock on T.S. Page B 3/4 3-2 states that:

"On decreasing reactor coolant loop temperature, P-12 automatically removes the arming signal from the steam dump system." This is not substantively consistent with Reference 5, Figure 7.2.1-1 which shows that'it is the arming signal for the condenser dump valves and atmospheric dump valves .hich is removed and then with the exce tion of 3 cooloown dump valves (to the condenser). The steam generator Power Operated [ atmos:heric] Relief Valves (SG 00RVs), are not affected: Please correct the Basis.

06/01/84 39 Revision A 1

O e A set point of 553-551*F is prov.ided. Provide the basis for this which should be consistent with our query under earlier Sec-tion 3/4.1.1. Boration Control concerning T.S. page 3/4 1-6, '

" Minimum Temperature For Criticality."

Item 10e. (Proposed).

To completa the list of ESFAS interlocks, it is necessary to add an item identified as 10e. Low T,yg.

The safety reasons for this are described under the later Item 11.b (Proposed) of this section.

Item 10c: Interlock, Reactor Trip, P-4. -

This currently reads as: " Reactor, Tri*p, P-4; with NA (Not Acol.icable) trip.setpoint & Allowable values." However, should not this item read as:

10c. P-4-with Trip Setpoint and Allowable values defined as in Reactor Trip to Table 2.2-1, with the uception of: " Power Range, Neutron Flux, Hign Negative Rate."

The basis for this is provided in Reference 5, Figure 7.2.1-1 (2 of 16),

Revision 42. The licensee should explain wny Reactor Trip Signals ini- ,

tiating P-4 include all items in Table 2.2-1 with the exception of " Power Range, Neutron Flux, High Negative Rate." The licensee shall evaluate and propose -

Item 11 Proposed: -

There is a need to add a new Functional Unit not addressed in the current T.S. , but wnien is a part of ESFAS. This is:

"Close Feecwater Isolation Valves & Close Feecwater Main & Bypass Moculating Valves." (See Referenca 5, Figure 7.2.1-1 (13 of 15)

Revision 34.)

This Functional Unit is initiated by:

a.

Reactor Trip P-4, & Low T,yg. -

b. Reactor Trip P-4, & Steam Generator Level - Hign Hign P-14,

c. Steam Generator Level - High High P-14 (see 5 above),

d. Safety Infection (see 5 above). "

Trip Set Points would be in accordance with the related values in earlier Items 10 and 5 of this section.

06/01/84 40 Revision A

Reference Item lib above, involvine Reactor Trip P-4 & Steam Generator Hich Nicn Level P-14 The NRC has observed potential situations of concern involving this interlock.

NRC Safety Concern A: A review of the logic of this interlock, Reference 7, Figure 7.2.1-1, (13 of 16), Revision 42 shows that if a SG-Hi Hi occurs, Turbine Trip, Trip of MFW Pumps, closure of MFW isolation and control valves occur, but the reactor is not tripped if the Nuclear Power Level is below P-8 (48% Power Level ), Reference 7, Figure 7.2.1-1, Revision 42, (18 of 18). This would then cause another occurrence which would be effectively a loss of main feedwater to the reactor at a nominal power level of 48%.

NRC Safety Concern B: The existing FSAR, Reference 7, Section 15.2.10.1, Revision 15, shows that a feeowater malfunction at . full power is not -

terminated by a neutron Flux Power trip, but by a SG-Hi Hi (i.e. , P-14) signal initiating Turbine Trip, Trip of MFW Pumps, Closure of MFW Isolation and MFW modulating valves. Turbine Trip will trip the reactor (if initial power level is above P-8). However, if the feedwater malfunction is ini-tiated at zero power FSAR, Reference 7, Section 15.2.10.2, "Results "

first paragraph, the consequences are a rapid increase in nuclear powee which will cause a reactor trip from the neutron flux l'ow power, 25%,

setcoint, and 35% (Limiting Safety Value in Analysis) and hence generate

{ a P-4 signal, but will not correct the initiating cause of the faulted 1 main feeawater contrel system until SG-Hf Hi level is subsequently ini- --

tiated and effects closure of MFW isolation valves. Whereas the FSAR evaluates the first event of this sequence by reference to the event of i " Uncontrolled Rod Cluster Control Assembly Bank Withdrawal From A Sub-critical Condition," the FSAR provides no evaluation of the subsequent

event including the ONBRs resulting from any restoration of reactivity

, before~SG-Hi Hi ultimately effectively closes MFW isolation valves. This latter event from zero power can also occur at any intermediate power

' level, with and without automatic rod control, and there is currently no analysis whien evaluates the worst case.

i NRC Safety Concern C: The licensee has provided no information on " Safety 3 Analysis Limits" that would be applicaole to Permissive P-8 in evaluating **

the acove events. If the' allowance is ultimately of the same order as for the Power Range, Neutron Flux - High and Low Set Point Trips, i.e., acprox.

+10 percentage point, then Safety Concerns A and 8 could be occurring at up to 58% power level.

In respect of NRC Safety Concerns A, B, and C above, we consider the pro-

! posed T.S. in respect of the related permissives and interlocks to be non-conservative with respect to Re,alatory Requirements. The licensee should review the safety consequences of each of these potential NRC concerns and respond with a safety evaluation with proposed changes to the T.S. as

appropriate. This could be considered a Generic Issue.

{ General: In view of the consequences of the bypass of reactor trip on l turcine trip below P-8 for the events protected by trip of turbine on 06/01/34 41 Revision A

r- - - _

Steam Generator Hi Hi., the licensee should review t,he analyses for all other Condition II through IV occurrences to cetermine wnether the con-clusions deriving from the existing evaluations need to be altered. This could be considered a Generic Issue.

Reference Item 11(a) above, involving Reactor Trip P-4 and Low T,yq, Reactor Trip P-4 together with Low-T,yg causes closure of the MFW isolation-valves and MFW Modulating (Control valves) thereby isolating the reactor from any faulted [on non faulted] feedwater system.

The safety significance of the parameter, Low T,yg, as expressed in the ..

FSAR derives (a) from its inclusion in the ESFAS under Reference 5, Figure 7.2.1-1, (13 of*16), Revision 34 and (b) a description in Reference 5, Section 7.7:1.7 under the title Steam Generator kater Lavel Control, in the following terms: *

" Continued delivery of feedwater to the steam generators is recuired as a sink for the heat stored and generated in the reactor following a reactor trip and a turbine trip. An override signal closes the feedwater-valves when the reactor coolant is below a given temoera-

.ture, and th'e reactor has tripped. Manual override of the feecwater control system is available at all times'."

This P-4/ Low T,yg combination does perfom a safety function in preventing _ _

excessive cooldown after the reactor is tripped, but has.never been incorporated, or discussed in the Section 15 FSAR analyses (Reference 7) for this purpose.

• Within the FSAR under Reference 7, Section 15.2.10.1 " Excessive HEAT REMOVAL QUE TO FEEDWATER SYSTEM MALFUNCTICNS" stata that:

"An accidental full opening of one feedwater control valve witn tne reactor at :ero power and the above sentioned assumptions, the maximum reactivity insertion rate is less than the maximum reactivity insertion rate analyzed in Subsection 15.2.1, Uncontrolled Control RCCA Bank Witharawal from a Suberitical Condition, and therefore. the results of the analyses are not presented. It should be notec that -

if the incident occurs with the unit just critical at no loac, tne reactor may be tripped by the power range high neutron flux trip (Iow setting) set at approximately 25 percent."

"For all excessive feedwater cases continuous addition of cold feec-water is prevented by closure of all feedwater control valves, a trip of the feedwater pumps, and closure of the feedwater pump d.. charge valves on steam generator high-level."

This event from :ero and higher power levels (already discussed under earlier Item lib) is initially protected by the high neutron fluxtrip; however whilst this provides immediate protection, the main feecwater is not isolated and continue to cooldown the reactor with continued reactivity addition. The licensee must confirm that acceptance criteria for tne reactor system are not exceeded if further protection must wait for Steam 06/01/84 42 Revision A

Generator Hi Hi Level to trip the MFW pumps, and together with existing Reactor Trip to provide Main Feedwater Isolation. Or, is it necessary to depend on an earlier " Isolation of Main Feedwater" from the combination of the existing reactor trip P-4 signal already provided and a related Low T,yg.

Inclusion of the P-4 and Low T,yg interlock into the T.S. would provide more reliability in protection for this event in conformance with the diversity critaria of 10 CFR 50 Appendix A, GDC Criterion 22 in support GDC 20. Without this, there is no diversity for protection from this continuing event. The proposed T.S. should require T,yg Low to be incor-potated into the T.S. to meet the above Regulatory Criteria. The licensee shall evaluate and propose.

I The licensee shall evaluate this issue with our concerns expressed under Table 3.3-4, Item 11 proposed Reference Ite'm 11(b) above, NRC Safety Concerns B and C'to which this is directly related. i The presence of Low T,yg, without T.S. cons,iderations of Set Point, Maximum Errors, Channel Reliability, Applicability MODES and Action Statements raises concerns about the consequences of a single failure.

For example, a failure low, remaining undetected, could comoine with a Reactor Trip from full power to close Matn Feedwater (containment] Isola-tion valves and Main Feedwater Modulatirig valves and cause a more severe transient than would otherwise be necessary. The Licensee snould evaluate tne consequences of single failure on aopropriate Conditions II, III, and

• l IV Occurences, and propose as necessary. .

. l Item: Reference 7, Section 15.2.14, page 15.2-38, Revision 43, which is the I Accident Analysis for " Inadvertent Operation of ECCS Ouring Power coeration," l states that l

Spurious ECCS operation at cower could be caused by ocerator erre or a false electrical actuating signal. Spurious actuation may be assumed to be caused by any of the following:

1

1. High Containment pressure

2. Low pressurizer pressure "j

3. High steam line differential pressure l

4. High steam line flow with either low average coolant temperature or low steam line pressure.

Please exclain the signals 3 and 4 since they do not accear in the TABLE 3.3-4 just reviewed, nor do they seem to appear in the Logic Diagrams of the Licensirg Basis in the FSAR to reference 5. The Licensee shall evaluate and prepose.

06/01/64 43 Revision A e

~ _ _ . . _ . _- . . _ _ _ _ .

Item": Reference 5, Figure 7.2.1-1 (2 of 16) Reactor Trip Signals The reference to Safety Injection Signal (Sheet 8) is inaccurate. This signal is from the ESFAS and not directly from the SI signal.

06/01/84 44 Revision A i-

)

<-n. -- . . . , - -,e- ,,,,,---~,----,n--,-,,,---r . - - -

TABLE 3.3-5 ENGINEERED SAFETY FEATURES RESPONSE TIMES 7a.I Item 22: :nie ni = of Saf ny injecu an by: c atainment pressure-Hign.

A value of 1 27 secs (without offsite power) is given.

Reference 5, page 7.3-8 shows that initiation time of ESFAS from this source is a maximum of 1 sec.

No events in Reference 7, Section 15, have been directly analyzed using this sensor as the prime initiator above the P-11 interlock although it is relied uoan for diverse protection. However, it is the only automatic initiation of Safety Injection protection below [P-11]. Other events deo.endent upon a SI generating signal, particularly circumstances descibed under items 34 and 4a below, shows safety analyses limits of $ 12 secs.

(with offsite power) and 1 22 secs (without off site power). ,

At this time, the proposed T.S. value is less conservative than others used in Safety Analysis. The licenses shall evaluate this difference and propose accordingly.

~ f 72 rtem 23: Initiation of " Reactor Trip (Frem 5:) by containment Pressure-Hign The descriptor (From SI), should be deleted as it is incorrect.

The response time is give is < 2 secs and tnis different from the FSAR,

~

Reference 5, page 7.3-8 which gives a maximum time of 1 sec.

This value is less conservative than the FSAR and the licensee sna11 evaluata and propose ac:ordingly.

Item 2c: "Feedwater Isolation" from Containment Pressure-Hign The response time is given as 5 9 secs.

Reference 5, page 7.3-9 shows that initiation of ESFAS frem this scur:e is a maximum of 1 sec.

Table 3.6.2 of the T.S. provides isolation times of < 5 secs for main feedwater containment isclation and < 10 secs for' main feedwater to Auxiliary Feecwater Isolation. A total time to isolation of MFW, fr'em Containment Pressure-High, of i 11 secs seems aopropriate to availacie equipment.

There would then be a conflict between the response time of ( 9 secs in the proposed T.S. and the potential uslue of up to 11 sec fr3m other licensing basis information.

No event in Reference 7, Section 15.1 through 4 uses this particular isolation in time Analyses. However, this is a imcortant factor for containment integrity during a Main Steam Line Sreak in containment. The value used as tna Safety Analysis Limit shall be provided by the licensee,

~~

06/01/84 45 Revision A l

l

~

comsared with proposed T.S. Item 2c and any differences evaluated, and T.S. proposed as acpropriate.

7C i Item 2d: Containment Isolation - Phase A, from Containment Pressure-Hign The proposed T.S. values are 18 I 3) (with offsite power) and 28(*) without offsita power.

Reference 5, page 7.3-8 shows that initiation of ESFAS from this source is 1 sec. .

Table 3.6-2 shows Maximum Isolation Times of up to 15 secs for Reactor Coolant Pressure Boundary Isolation valves. A minimum total time to containment and isolation (for the RCPB] of 16 sacs seems feasible, plus 10 secs giving 25 secs total without offsite power.

The proposed T.S. values should be check 3d aga' inst those used as Safety Analysis limits for related Conditions II, III, and IV occurrences using SI. Values used by licensee shall be provided, comcared with Item 2d.

and any differences evaluated.

74 Item 2e: Containment Purge and Exnaust Isolation, from Containment Pressure-High This is given as N.A. This is not so; response times have be used to minimize offsite consequences of any Condition occurring whilst contain-ment purge & exhaust is being used. This proposed T.S. is less conserva-

} tive than the licensing basis. The licensee shall evaluate & propose.

Item 2f: Initiation of Auxiliary Feedwater from Containment Pressure-Hign.

N The licensee proposes N.A. but earlier review shows AFW initiation on Containment Pressure-High and especially in MODES 3 and 4 This is less conservative than :ne licensing basis; the licensee snail t evaluata and propose. _

Item 2g: Initiation of Nuclear Service Water (NSW) from Containment  !

Pressure-Hign .

This response time is given as $, 65(3)/76(4) secs. i The superscript 3 does not seem appropriate; whilst the related Notation on T.S. Page 3/4 3-33 refers to aesence of diesel delay (i.e., no loss of offsite power), it describes start uo of ECCS equipment but does not include tne requirement for " Isolation and Startup of Nuclear Servica Water Pumps as described in Functional Unit 1 of Tacles 3.3-3 and 3..s-4.

The same comment applies to superscript 4 which applies to the circum-stances without offsite power. The licensee should propose an accurate description of these circumstances; the current description does not meet the intent. ,

06/01/84 46 Revision A

Reference 5, page 7.3-8 shows that initiation of ESFAS from this source is 1 sec.

No other information is available on Safety Analysis Limits because, contrary to Regulatory Requirements, this value has not been used in the Safety Analysis of the FSAR in respect of AFW sucplies. In other sec-  !

tions of this review, the licensee has been asked to re-evaluate Safety i Analyses to recognize this fact. Parallel with this, the licensee shall 3 8

identify the Actual Safety Analysis Limit to be used for this response, compare with the proposed T.S., and repropose as appropriate. Any occur-rences required to utilize Nuclear Service Water must be considered non-7 conservative with respect to these values currently presented in the FSAR to Reference 7, Section 15.

Item 2h: Initiation of Component Cooling Water from Containment Pressure-High This response time is gi n as 65(3)(3)/76(4)(2) secs.

ll The description of suggescript 2 under Table Notation on T.S. Page 3/4 3-33 l is incomplete. The licensee shall propose an accurate description of these circumstances including its dependence on Nuclear Service Water; the licensee snould confirm that this cooling water supply information is for this safety related service. -

. I i

~ '

Reference 5, page 73-8 shows the initiation of ESFAS from this source is 1 sec.

No other information is available on Safety Analysis Limits used in the l.-

l i

FSAR. The licensee shall provide this information for related Condi-tions II, III, and IV Occurrences for both on-site and offsite power. This l

i.

information shall e evaluated and the licensee shall propose. At this time, considering the non-conservative circumstance with NSW AFW supply, it must be presumed that any Occurrence recuired to utilize tne Nuclear Service Water must be considered non-conservative with resoect to tne values currently presented in the FSAR, Reference 7, Section 15.

Item 21: " Start Diesel Generators" from Containment Pressure-Hign A response time of i 11 secs is given. -

t Reference 5, page 7.3-8 shows that initiation of ESFAS from the source I is a maximum of 1 sec.

l No evaluation in Reference 7, uses this sensor as the prime initiator above the P-11 Interlock, although it is relied upon for protection acove, and directly for protection below (P-11]. Other events dependent upon a SI generating signal particularly, :.ams 3a & 4a below, show safety analysis limits of i 10 secs for this value.

In respect of current safety analyses limits, therefore, it acpears nat the proposed value is less conservative than the Safety Analysis Limits.

The licensee sna11 evaluate and propose.

06/01/84 47 Revision A

- __ _ u- -

We note that Reference 5, page 8.3-6, describes testing of diesels on 11 second starts and if initiating times of 1 and 2 seconds were allowed for, this would mean actual times of 12 and 13 secs from the initiating signal. The licensee shall clarify, evaluate and propose.

Item 3: Pressurizar Prussure-Low This title should be modified to read as Pressurizer Pressure-Low (Safety injection) as Pressurizer Pressure-Low Is a Reactor Trip only.

The initiation time of all ESFAS Functions from this son'sor is < 1 sec (Reference 5, page 7.3-8). This is also the same initiation tiEe for Containment Pressure-High. Since both or either of these initiators can be available in Occurrences involving SI, and initiation times are the same, our comments and conclusions under earlier Item 2 can be directly -

referenced for items under Item 3 in cases wnere the, proposec response-

- time is the same for e given ESFAS function. .

7hfItem3(a): " Safety Injection (ECCS)" on Pressurizer Pressure-Low [S!]

Values of 1 27b)/12(3) secs are proposed.

Reference 5, page 713-8, shows a maximum initiating time of ESFAS 1.0 secs I for this signal. -

The value of 12 secs (with offsite power) is consistent with safety analysis limits given for the MSLB in reference 7, page 15.4-10, Section 7 where "In 12 seconds, the valves are assumed to be in their-final position and pumps are assumed to be at full speed." For.the other case with Loss of Offsite Power (LOOP) "an. additional 10 secs delay is assumed to start the diesels and to load the necessary equipment onto them." Further, this particular analysis appears to initiate the event on Pressure Pressure-Low (SI).

The proposed value of i 12 secs appears witnin the licensing Dasis of 12 secs.

) The proposed value of 27 secs (with LOOP) is hcwever larger tnan :ne value T of 22 seconcs from the reference cascribed above (i.e., 12 secs + 10 secs "

delay for start of diesel). This value of 27 secs therefore accears less conservative than the FSAR, reference 7, page 15.4-10, anc the licensee shall evaluate and propose.

Item 3b: "Reactnr Trip (from SI)" on Pressurizer Pressure Low (SI]

7g The descriptor (from SI) is incorrect and should be deleted.

A value of < 2 secs is proposed. The FSAR in Reference 5, page 7.3-9 ouotes a vaTue or 1 1 secs.

The proposed T.S. value appears less conservative than the Safety Analysis Limit and the Itcensee should evaluate and propose.

06/01/84 48 Revision A

. - . -. -- -- - - - . _ . =

Item 3c: "Feedwater Isolation" From Pressurizer Pressure-Low (SI)

The proposed T.S. is 1 9 secs.

Reference our comments and requirements under 2.c. above. l

] M Item 3d: " Containment Isolation - Phase A" from Pressurizer Pressure-L:w (SI)

I The proposed T.S. is 1 18(3)/28(4) secs.

Reference our comments and rea'uirements under 2.d. above. ,

?

I Item 3e: " Containment Purge & Exhaust Isolation" From Pressurizer K Pressure-Low (SI)

, {

The proposed T.S. is NA. -

Reference our comments and requirements under 2.e. above.

7j Item 3f: "Auxi n ary reeawater" Initiation by pressurizer Pressure-i.ow (SI)

The licensee proposes NA (not applicable).

Safety injection logic closes the main feeowater isolation valves for every event in which SI is initiated (reference earlier sections of this review Table 3.3-4, proposed item c). Therefore, every such event -

initiated by a SI initiator must be analyzed with a restoration of AFW f and a related response. time. ,

It is outside the licensing basis, not to a procose a value for this c'esconse time. This T.S. value is therefore non-conservative; the licensee shall evaluate and propose.

Item 2g: " Nuclear Service Water System" Initiation from Pressurizer Pressure-Low SI The T.S. value is given as 76(1)/65I 3) secs.

Our comments on 65(3) are as for our earlier 2g. ,

- With respect to superscript (1) on 76; wny is this cifferent to Containment Pressure High which is 76(3) when the concomitant SI signal generates the same equipment requirements. Superscript (1) now provides for Si and RHR pumps wnereas (3) did not. Also, superscript (1) , if it is to Da used should include Isolation and Staat of Nuclear Service Water System (NSW).

Reference our comments and requirements under earlier 2g.

Item 3: General The licensee is to evaluate each of his superscripts (1) , U , (3) ano (4) and ensure that they are complete, accurate and consistant with all tne related ESFAS initiating signals and functions.

06/01/84 49 Revision A

,p -- , - _ . _ _

• e This position appears inaccurate & confusing to the extent that it must be considered non-conservative.

Item 3h: Initiation of Component Cooling Water frem Pressuri:er Pressure-Lew (SI)

The proposed T.S. is 5 76(3)/65(2)(a) secs.

See our comments and requirements under 2h. and 3. General above.

Item 3f: Start Diesel Generators from Pressurizer Pressure-Low (SI) .

The,T.S. value is i 11 secs. -

See our comments under 21. above which are substantively appitcable to this item. Therefore, the preposed item is .less conservative than the safety analysis limits; the licensee shall evaluate and propose.

Item 4: Steam Line Pressure-Low The initiation time for all ESFAS functions for this sensor is given as

> 2.0 see in Reference 5, page 7.3-8. This compares with only 1 sec for item 2, Containment Pressure-High and Item 3, Pressurizer Pressure-Low (SI). Since again, all these 3 initiators can be available in occurrences involving SI, our comments and conclusions under 2 and 3 can be referenced with the condition that actual response times under item 4 could be 1 sec longer. We note however, that functional response times specified under '

4 remain the same (in general) as under Items 3 and 2 and do not apparently -

provide for this differential of 1 sec. The licensa shall evaluate and propose.

Item da: " Safety Injection (ECCS)" Initiation on Steam Line Pressure-Low I

These values of i 12 3)/22(4) agree with the Safety Analysis Limits of the Licensing Basis FSAR.

Item ab: " Reactor Trip (Frem SI)" from Steam Line Pressure-Low. '

t-Thedescriptjeg(fromSI)isincorrectandsnouldbedeleted. .

This value of 12 sees agrus with Reference 5, page 7.3-8. -

g Item 4c: "Feedwater Isolation" from Steam Line Pressure-Low The proposed T.S. is 1 9 secs.

Reference our comment and requirements under 2c. above modified by the fact that there appears to be a larger conflict between the response time of 19 secs and the potential value of up to 11 + 1 = 12 seconds from Licensing Basis Information.

06/01/84 -

50 Revision A

i Item 4d: " Containment Isolation - Phase A" on Steam Line Pressure-Law The proposed T.S. is i 18(3)/28(4) secs.

Reference our comments and requirements under 2d. aoove, modified in that proposed T.S. times appear faasible with the additional delay of 1 sec.

7 W Item 4e: " Containment Purge and Exhaust Isolation" on Steam Line Pressure-Low

\

5 The proposed T.S. is NA. ,

Reference our comments and requirements under item 2d. above. J Item 4f: " Auxiliary Feedwater Pumps" initiated by Steam Line Pressure-Low I t

The proposed T.S. is NA. ,

Reference our comments and requirements under 3f. above.

g item 4g: " Nuclear Service Water" initiated on Steam Line Pressure-Low The proposed T.S. is 1 65(3)/76(4) secs.

Reference our comments, requirements, and remarks under 2g., 3g., and 3 General above'.

1 7]. Item ah: Steam Line Isolation on Steam Line Pressure-Low.

The proposed TS value is 1 9 secs.

9eference 5, page 7.3-8 states that the maximum allowable times for generating steam break protection are (1) from steam line pressure rate, 2 secs, and (2) from steam line pressure-low 2 secs. Further, Refer-ence 7, page 15.4-6 states that the fast acting steam line stoo valves are acesigned so close in 5 secs...". A minimum closure of 7 secs seems i likely, For actual safety analysis limits, Reference 7. Table 15.4-1 (1 of 4) anc 15.4-1 (2 of 4) both show a difference of seven (7) secs cetween arriving ";

at the " Low Steam Line Pressurs Setpoint" and "All main Steamline Isolatten I Valves Closed." [In the case of Feedwater System Pipe Ruoture]

I The proposed TS value of 1 9 secs is therefore greater than the Safety Analysis Limit.

The proposed TS must therefore be consider 1 less conservative for this event. The licensee shall et.luate and propose. j Item 41: " Component Cooling Water" Initiation by Steam Line Pressure-Low Prooosed T.S. value is 65(a)(a)/76(3)(4) .

Reference our earlier comments and requirements under 2h and 3h. acove.

51 Revision A 06/01/84

o . - . . _ _ __ ..

Item 4j: " Start Diesel Generators" by Steam Line Pressure-Low.

Proposed T.S. value is 1 11 secs.

Reference our comments and requirements under 21 above.

Item Sa: " Containment Spray" - Initiated on Containment Pressure-High-Hign Licensee shall provide the Safety Analysis Limit and compare with the proposed value of 1 45 secs. Evaluate and propose as necessary.

Item Sb: Containment Isolation - Phase B on Containment Pressure-High-Hign This is' proposed as Not-Applicable. The Ifcensee should propose why this is so when it appears that TS Table 3.6-2 Containment Isolation valves.

• Maximum Isolation. Time (secs), applies oni.y to closure from receipt of signal, and may'not include'the ESFAS Response Time. Reference especially T.S. s ge 3/4 6-30 wnere main steam line isolation is specified at 5 secs compred eith the same value quoted on Reference 7, page 15.4-6 wnich

-states that these fast acting steam line valves are designed to close in 5 sees and Safety Analysis Limits have been shown as 7 secs under Item 4h.

above.

What b needed to supplement the information in T.S. Table 3.6-2 is the -

i ESPAS response time as defined in Reference 5, page 7.3-7, Revision 36, and which values are quoted at 1.0 sec for initiation from containment I pressure (relatec page 7.3-7), and also as 1 sec for closing main steam line stop valves on Containment Pressure-High (Vigh]. It ar. pears this item should read as:

Sb. ESFAS Input to containment Isolation - T+ase 8 - 1 sec The licensee shall clarify, identify the related Safety Analysis Limits, and evaluate as accrocriate. Until then, the Preposec T.S. must te consicered non-conservative witn respect to the Licensing 3 asis.

Item Sc: Steam Line Isolation on Containment Pressure Hign-Hign The pretosed T.S. value is 1 9 secs. ,

Referenca 5, page 3.7-8 shows centainment pressure initiating ESFAS signals with a 1 1 response time. Item 4h. above snows fast acting st.co valves  !

closing in 5 secs. giving a total time of 1 6 secs.

Since MSIV actuation under Containment-Hi Hf can be caused by MSL3 wnich provides for a maximum of 7 secs above, the proposed value of 9 secs appears less conservative. ,

A comoarison also with values used in assessing environmental releases from containment should also be made.

06/01/84 52 Revision A

--- - ,m. m . ..._ _ _ _ . . . _ . _ _ . . _ _ _ . . _ . _ __ , _,_

t The licensee shall identify the Safety Analysis Limits used for this Steam =

Line Isolation, including the MSLB in containment, evaluate against the proposed T.S. value and propose as appropriate. Until such time, the  ;

current value appears non-conservative. .

I Item 6a: Turoine Trip on Steam Generator Water Level-High High The proposed T.S. is NA, i.e., not applicable. ,

Reference the licensee to our comments under Table 3.3-2, Item 16 where it is shown that it is used within the Licensing Basis. ,

The proposed position is non-conservative with respect to the Licensing Basis. The licensee shall evaluate and propose in accordance with our .

review under Table 3.3-2, Item 16.

7m[ Item 6b: "Feeawater I' solation" Initiated by Steam Generator Water '. -

'l

• Level-High High I

The proposed T.S. is i 13 secs.

Reference 7, Table 15.1.3-1 shows that "High Steam Generator level trip of the feedwater pumps and closure of feedwater system valves, and turbine trip" is based on an ESFAS time delay of 2.0 seconds.

Table 3.6.2 of the T.S. provides isolation times of < 5 secs for main feedwater containment isolation and <~ 10 secs for maTn feedwater to Auxiliary Feedwater Isolation.

A total time to isolation of MFW of i 13 secs seems appropriate to avail-able equipment.

However'the current safety analysis depencing on this response time is f that for the Excessive Cooldown occurrence under Reference 7, page 15.2-29, and for this, no value is quoted for isolation of main feecwatar wnich is the initiator of the event. However, Figure 15.2.10-2 shows that with inf-f tiation of the event caused by one faulty control valve, it takes 32 secs to reach the SG-High-High Level with a mass increase of 35% of initial, and thereafter does not increase further. This implies zero closure time.

Since it it expected to take another 13 secs to actually isolate, we could -

assume an additional mass increase of another 13% to give a total of approx. 1.48 the initial value.

The above additional Main Feedwater level can affect the consequences of the event at power, if there has seen a trip, with a potential for power restoration and/or overfill of the S-G to cause water ingress into the main steam lines. Additional h , it can have consequences of potentially larger importance for the event occurring from zero suncritical power.

Reference also our concerns under item Table 3.3-4, item lib and 11a acove.

The licensee shall evaluate the related concerns, including the extenced MFW valve isolation times, to determine their safety significance, and 06/01/84 53 Revisien A

.. . . . .. - . - - . . . ~ . . . . .- . . . . - - . . .- .

rpt (c. b_ [

propose as required. Until that time, it must be concluded that sirce a zero (0) value has been used in the current analysis, that the licensee has a potentially non-conservative situation with respect to Regulatory Recuirements of Reactivity Control and Regulatory Concerns for Flooding of the Main Steam Lines. J Item 7a: " Motor-Oriven Auxiliary Feedwater Pumps" initiated by SG Level-Low Low Item 7b: " Turbine-Oriven Auxiliary Feedwater Pumps" initiated by SG Level-Low Low Proposed T.S. response times are given as < 60 secs. ,

The FSAR Safety Analysis Limit is 51 secs; Reference 7, Table 15.4-1 (1 of 4) and 15.4-2 (2 of 4) where the difference between SG Low-Low and auxiliary feecwater delivered to steam generators is 51 secs. The current proposed T.S. value is therefore conservative witn respect to, the current 1 safety analysis limit. -

However, the current safety analysis limit of 61 secs currently used appears to be a sistake and not in accordance with Regulatory requireunts.

The only safety related water source available for Auxiliary Feedwater, is the Nuclear Service Water System.

Reference 22, page 10.4-14a, states that "All three AFS pumps are normally supplied from a common leader which can be aligned to the upper surge tank, the auxiliary condensate storage tank, or the condenser hotwell. Each of --

these sources are provided with motor operated valves with control room operation. The assured AFS pump suction is from the Nuclear Service Water System. The A motor drive is aligned to the A NSWS header and the B motor driven pump is aligned to the 3 NSWS header. The turbine driven pump is aligned to both channels. Each source is provided with diesel aligned motor operated valves ~which open automatically on how suction pressure" (with a proposed T.S. response time nf 13 secs].

Earlier information .under this T.S. Tabie 3.3-5 st'cws the.t the response time for Nuclear Service Water Sucoly is 65 secs, assuming offsite power availaole and 76 secs assuming loss of offsite power whereas the Safety Analysis Limit used in the FSAR is only 61 secs. On this basis, all '

Conditions II, III, and IV 6ccurrences involvirig AFW supply would need to be re-evaluated to establish acceptability.

l The NRC does notice from Reference 5, Table 8.1.2.1 entitled " Maximum Loads to be supplied from one of the Redundant Essential Auxiliary Power Systems" that the related loading sequences for pumping equipment, alone, might enable an earlier response time then given in Table 3.3-5. e.g.,

Nuclear Service Water Pumps can be available 35 secs and AFW, 4b secs, after Blackout or LOCA signal (further, the Table notation of Table 3.3-5 is inadequate to clarify the position]. '

The licensee shall clarify the available response time for AFW supply from the Safety Related Nuclear Service Water system, and include the conse-quences of additional delays due to inadequate suction pressure under 06/01/84 54 Revision A

- _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ = -

--.z- s__ a . _ ,. a- . - _ . . - - - . . .

Item 11, below. If this is confirmed at from 65 to 70 secs, or any longer time than used as the existing Safety Analysis Limit in the FSAR, then acce:: table re-evaluation of all Conditions II, III, and IV occurrences involving AFW supply, are required by 10 CFR 50.36.

Our current evaluation is that the response times in the proposed T.S.

are non-conservative in respect of Regulatory requirements.

, Item 8: " Steam Line Isolation" on Negative Steam Line Pressure Rate-High Proposed T.S. value is 5 9 sec.

Reference 5, page 7.3-8 states that the maximum allowable time for generating the ESFAS MSIV isolation signal from a Steam Line Pressure Rate circumstance is 2 secs, tne same as for item 4h. above.

Our commenhs and requirements therefore are the same as under item 4h. ,

We appreciate that this signal is generated at below P-ll, but with the existing proposed Saration Control T.S. we must continue to evaluate this value as non-conservative.

The proposed T.S. value is greater than the Safety Analysis Limit of seven (7) secs and must be considered less conservat.ive for this event: The licensee must evaluate this difference and propose.

Item 11: " Automatic Re-alignment of AFW Supply on Low Suction Line Pressure" l

[The existing description should be changed to more accurately state this action].

Proposed T.S. value is 13 secs.

Note our comments under 7a. and 7b. above. Although this response time may

,. be in accorcance with current plant engineering, it is not in accorcance with the existing 3afety Aralysis Limit for Auxiliary Feedwater Scooly wnich, on current information, has pre succesed no sucn transfer time.

If a tank has been lost because of seismic action, we cannot assume a residual 15 secs supply at this time. ,

At this time, until the evaluation of 7a. and 7b. above is completed, we must evaluate this delay as non-conservative with respect to currently used Safety Analysis Limits which in themselves are non-conservative with  !

respect to Regulatory requirements.

The licensee will evaluate and propose.

7, ,

Item 12: " Automatic Switchover to Rec.irculation" on Low RWST Level

,. . Response time proposed as 1 60 secs k The licensee shall provice the bases for this value and evaluate against this 1 60 secs, and propose as necessary.

J

. ~

2 06/01/84 55 Revision A

t Item 13: Station Blackout Item 13: General The Licensing Basis FSAR, reference 6, page 9.2-10 describes-how station blackout causes startup of all Emergency diesel generators and alignment of [NSWS and CCW]. Why is this not included under this item 13 " Station Blackout."

The Licensing Basis FSAR, reference 7, Section 15.2.9 under LOSS OF QFF-SITE POWER TO THE STATION AUXILIARIES describes a set of Protection Actions for the plant, all which have related response times. Why is this information not provided under this heading?

The. absence of most of the information on Functional Units and Related -

Response. times required to protect the facility on Station Blackout conci-tions makes the proposed T.S. non-conservative with respect to tne Licensing Basis. The Licensee shall evaluate and propose.

Item 13a: " Start Motor-Driven AFW Pumps" on Station Blackout Item 13b: " Start Turbine-Oriven AFW Pumps" on Station Blackout Proposed T.S. response times are 1 60 secs.

Reference our comment under 7a. and 7b. above. -

These values are non-conservative with respect to Regulatory requirements and the licensee shall evaluate and propose.

Item 14: " Start Motor-Driven Auxiliary Feodwater Pumps" on Trip of Main Feedwater Pumps Proposed T.S. value is < 60 secs.

Reference our comments under 7a. and 7b. acove together with the necassity for licensee action.

At this time, these values are non-conservative with respect to regulatory ,..

requirements, and the licensee shall evaluate and propose. -

Item 15: Loss of Power: "4 Kv Emergency Bus Undervoltage-Grid Degraded Voltage."

Proposed T.S. response time of i 11 secs.

Reference our comments under T.S. Table 3.3-3 Item 9 and Table 3.. 4 Item 9 and provide appropriate clarification.

No evaluation is possible at this time.

~~

06/01/84 S6 Revision A

i .

Item 15: Loss of Power Ites 15: General Our review comments under item 13 " Station Blackout" are fully apolicaole i to this item with the related conclusion that: 1 i

The absence of most of the information on Functional Units and related i Response Times required to Protect the Facility on Loss of Power makes I the proposed T.S. non-conservative with respect to the Licensing Basis. ,

The Licensee shall evaluate and propose.

Item [ Foot] Notw: Response time for Motor-Oriven Auxiliary Feedwater Pump -

Starts on All SI signals.

6 This is proposed as < 50 secs. - . s i

Reference our earlier comments for its inclusion in Items 2f. , 2f. , and j

4f. above together with the necessary Licensee Actions. ,

Reference our earlier comments under 7a. and 7b. above together with the  ;

necessity for licensee action.  ;

At this time, t,hese values are non-conservative with respect to Regulatory requirements and the licensee must evaluate and propose.

Item: Table 3.3-5, TABLE NOTATION on T.S. Page 3/4 3-33 ._

l These notations 1, 2, 3, and 4 must be expanded to include Component Cooling Water System Isolation and Pumps, Nuclear Service Water. System (NSWS) Isolation & Pumps, and AFW re-alignment to NSWS and alternate sources as necessary. This will also enable verifiable consistency with the Notations used in the table.

See our comment under items 2g., Zh., 3g., 3h., 4g., and 41. accve.

Notation 2 of this Table states that:

(2) Valves 1KC3053 and 1KC3158 for Unit 1 and Valves 2KC305B and 2KC3153 for Unit 2 are exceptions to the response times listed in the table. The ,.

following response times in seconds are the required values for these valves for the initiating signal and function indicated:

)

2. b 3.b <

< 30 30((3)/40(#)

4.b 530(3)/40I#)

Since the functions 2b. 3b and 4b are all Reactor Trip functions, please explain.

Since these descriptors are apparently incorrect, provide the correct descriptors. ,

57 Revision A 06/01/S4 ,-

_ =- _ _ _

~ _ _ . . _ . . _ _ _ _ _ . . - _ _ _ _ _ _ _ _ . . _ __ _ ._.

Since sucercripts (3) and (4) used above make no mention of Comconent Cooling Water, [from which the valves derive] what do they mean?

What is meant by the Statement that the valves specified are exceptions to the response times listed in the Table. How do they affect the response times - do they increase, or decrease them, or have no effect. If they increase response time, by how much and what is the effect on the Actual overall response time, and has this been incorporated into the Safety Analysis of the Licensing Bas'is.

The Licensee shall clarify, evaluate and propose. Lack of accurate information on response times must be considered as non-conservative.

h

, =

06/01/84 58 Revision A l - . . . _ . --

-. ~ : . . : . _ . . _u  ;  ; . _

Section 3/4.4 REACTOR COOLANT SYSTEM, Section 3/4.4.1 REACTOR COOLANT LOOPS AND CCCLANT CIRCULATICN Item: GENERAL G.1 INTRODUCTION I

Concerning RCS Operability requirements, in MODE 3-5:

Werefertoourearlierdiscussionsblicenseerequirements-andespecially -

under Section 3/4.1.1, T.S. Page 3/4 1-1, 2 & 2a on Scration Control, T.S.,  !*

Page 3/4 1-20 & 1-21 concerning SHUTDOWN AND CONTROL ROD INSERTION LIMITS and TABLE 3.3-1 REACTOR TRIP SYSTEM INSTRUMENTATION generaT1y, including more ,

particularly items 2-21 (selectad) and items 12, la,15 and 21.

Under ou'r item T.S. TABLE 3.3-1; items 2, 5 & 6 et al, the licensee nas been-required to " Provide an anlaysis and evaluation of the consequences of Appli-cable Condition II, III and IV Occurrences, in MODES 3 througn 5, for an appropriate set of Technical Specification requirements to ensure Conformance to Acceptable Regulatory Criteria, and from this establish an appropriate range of Reactor Trip System Instrume1tation to Safety Related Requirements. This  !

evaluation shall be undertaken in conjunction with our concerns for current ,

technical specifications under section 3/4.4.1 REACTOR COOLANT LOOPS' AND COOLANT - - .

CIRCULATION of this review.

As part of this review, and as a safety justification for our concerns, we require inclusion of the following Occurrences and Considerations in the program, and as early determinants of our proposals in respect of RCS Loop Operability requirements in MODES 3, 4 and 5 (with loops filled).

G.2 OISCUSSION Item: CONSIDERATICN l

A numoer of factors determine our concern:

G.2.1 The increasea baron concentration discussed under Section 2/4.1.1 of -i this review.

l G.2.1.1 Increases shut down margin at temperatures above 200*F, and therecy l reduces the severity of any occurrences giving a return to power, l but only after reactor trip. Further the T.S. proposed by the licensee j does not include the increased baron concentration and RCS Operability i requirements are judged against those circumstances.

G . 2.1. 2 Because increased shutdown margins are available, in MODES 3, A and <

5, the licensee may now increase the level of withdrawal of all i movable control assemblies and still remain within the unchanged T.S.

condition of the allowable reactivity condition, keff of < 0.99.

Consequently, it coes not benefit those Occurrences initiated by fast positive reactivity excursions in which maximum power levels ulti-mately reached are substantively determined by given Response Times

)

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. to Trip. Further, events giving a return to power after reactor trip do not have improved initial protection; the reactor must still be 4

. tripped prior to effecting the increased shut down margin, and the elimination of virtually all " Safety Relatec" levels of neutron flux trip protection in TABLE 3.3-1 removes all current confidence in "available" Reactor Trips on Neutron Power; the only Safety Related Neutron Flux Trip from zero power subcritical conditions is the Power Range Neutron Flux Low Set Point and the proposed T.S. removes this from operability in MODES 3, 4 an.d 5. Further it has a Safety Analysis Limit' of 35% power (25% Set Point) and together with related high peaking flux factors under these conditions is sufficient to require all 4 RCPs running to ensure A.C.S. Safety in at least MCDE 3.

G.2.1.3 The increased baron concentrations give lees negative and more posi-

-tive mooerste coefficients wnica changes the complexi.on and natur.e of .

expected responses from " Licensing Sases Events." Under these cir- -

cuestances, it may not be possible to validly deduce the resulting responses and consequences without related analyses.

G. 2.1. 4 At this time we see no protection against positive temperature coefficients in MODE 3 [4, 5 & 6]. Proposed T.S. page 3/4 1-4

. concerning MODERATOR TEMPERATURE COEFFICIENT requires only that:

"the moderate temperature coefficient (MTC) shall be:

3.1.1.3.b. Less negative than - 4.1 delta k/k *F for .

all the rods withdrawn, and of cycle life (EOL), RATED -

THERMAL POWER condition." The T.S. proposes that this -

is " Applicable to MODES 1, 2 and 3" only. The licensee should also clarify this T.S. requirement which is apparently in error and applicable to MODES 1 & 2 only because of the " RATED THERMAL PCWER Condition."

G.2.2 Removal of ocerability requirements for all safety related reactor trips (exceot SI) in Modes 3, 4 and 5, has placed the reactor in nonconformance with the requirements of 10 CFR Appenoix A GCC 20,

" Protection System Functions" and GDC 22, " Protection System Indeoendence For All Occurrences Not Inititating Safety Injection."

j Further, only a limited number of automatic t' rips (6) are clocked by existing plant permissive. P-7, 2 are blocked by P-8. This leaves i an additional 9 from which automatic protection can potentially ::e previoed and which have been removed by unique action of tne T.S.

without any Safety Evaluation.

The proposed T.S. are nonconservative with resoect to Relulatory Requirements. They are also conconservative in respect to cne Licensing Basis. The Licensee shall evaluate and propose. ,

G.2.3 In MODE 3, down to P-11, for events initiating Safety Injection, the i engineering within the existing Licensing Basis, mignt allow 10 CFR 50 Appendix A GDC 20 and 22 to be satisfied in respect to reactor trip l

i and diversity. However, the proposed T.S. does not propose  ;

I 06/01/84 60 Reve sion A

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~.

operability of Reactor Trip from SI in this mode and offers no Safety Evaluation for the proposed change. Reference our review under Table 3.3-1, Item 17.

The proposed T.S. is not in conformance with the Licensing Basis, and is nonconservative. The licensee shall evaluate and propose.

G.2.4 In MODE 3, from P-11, to MODE 5, for events initiating SI, the plant is engineered and can be operated so that only one automatic trip of the reactor may be availacle; that from containment pressure-high.

On the above bases, plant engineering and operations would not be in conformity with regulatory requirements. The Licensee shall evaluate and propose. .

^

It may be possible for the plant to be operated in a manner to conform by not manually blocking the Main Steam'Line Pressure-Low Trip (at P-11] but constraining this blockage to a point at which SG pressure during cooldown is within an acceptable error band of the related Set Point Value. Under these circumstances, two (2) diverse automatic protections on reactor trip may be available.

In addition the proposed T.S.s do not require operability of the ,

Reactor Trip /ESF channel in this phas's of operations below MODE 3

[at P-11], to MCOE 4 even thcugh this is engineered into the Facility. No Safety Evaluation of this omission is provided. The FSAR assumes Safety Injection Protection in MODES 3 and 4. The proposed T.S. is not in accord with the Licensing Basis and is nonconservative. The. Licensee shall evaluate and propose.

G.2.5 ' Diversity of Safety Injection to the maximum extent for related Accident Circumstances can only be retained within existing plant engineering by recuiring that manual block of the Steam Line Pressure-Low ce celayed until SG pressures are within an aporopriate error cand of the Steam Line Pressure-Low Set Point. This could te down to a temperature of approximately 485-490*F in the RCS which would be in MCOE 3 before 1000 psig/425'F. (485-490*F is tne satur-ation temperature ecuivalent to 565 psig + 30 psig [cnannel error]

1.e., approximately 595 psig in the SG. . -

The licensee shall evaluate and propose.

I~ G.2.6 EVENTS OF CONCERN (A LIMITED SELECTION)

{[dL G.2.6.1 OCCURRENCES WITH RAPID REACTIVITY INCREASE Concerning " Uncontrolled Red Cluster Control Assembly Bank Withdrawal from SuD-Critical Condition."

Current Oceketed Analysis in reference 7, section 15.2.1, page 15.2-2 is based on four coerating locos. This event is possible down to and including Mode 5.

Current FSAR analysis trips the reactor on Power Range, Neutron Flux-Low Set j 06/01/94 61 Revision A

T*(F Point (25%) at a Safety Analysis Limit of 35% (reference page 15.2-3, item 3).

The principal determinant of ultimate power level is Doppler coefficient; contribution of moderator reactivity coefficient is negligible (reference page 15.2-3, items 1 & 2). The event is initiatad from not zero power (reference 7, page 15.2-4 item 3). 4 RCS pumps are operating.

Given the circumstances of the proposed T.S. , any T.S. allowing OPERABILITY of less than 4 RCS Loop in MCDE 3 would be in nonconformance with the current FSAR in a nonconservative manner, and the licensee would be required to evaluate and propose.

Furthemore; increased baron concentrations would not change this requiremerrt.

Additional events of a s'imilar nature, with a rapid increase in reactivity include:

a) Uncontrolleo Boron Oilution (reference 7, pages 15.2-13) b) Startup of an Inactive Reactor Coolant Loop (reference 7, page 15.2-19, revision 7) c) Excessive Heat Removal Oue to Feedwater System Malfunction (reference 7, page 15.2-30, revision 7) concerning initiation with the reactor at zero power). Until the licensee clarifies availability of MFW during MODES 3 through 5, this must be considered a potential occurrence.

d) Single rod cluster control assembly withdrawal (reference 7, Page 15.3-9, -

revision 7). Although the Licensing Basis is at 100% power, the cir-cumstances from zero power should be reviewed.

e) Ru::ture of a Control Rod Drive Mechanism Housing, at Zero Power (ref-erence 7, Page 15.4-30; revision 42).

f) Major Rupture of a Main Steam Line (see below).

yh ' G.2.6.2 STEAM LINE BREAKS: CCOURRENC25 Concerning " Major Rupture of a Main Steamline" This event is discussed in Accident Analyses in Reference 7, section 15.4.2 ano l Reference 8 item 212.75 page Q 212-47d & e, item 25. Reference 8 precosas that the resulting impact on shutdown margins from this event curing MCDES 3, 4 ano 5 are imoroved over that of the design basis (of zero power, just critical, Tavg - 557') as:

l "Ocerating Instructions require that the baron concentration be

! increased to at least the cold shutdown boron concentration before cooldown is initiated. This requirement insures a minimum of 1% .tk/k shutdown margin at a Reactor Coolant System temperature of 200*F. This condition assures that the minimum shutdewn margin experienced during the streamline ructure from zero power shown in the safety analysis is less than the case where safety injection l

J l -

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%(C actuation is manually blocked on low steamline pressure and low pressurizer pressure." .

i This position gives no measure of the resulting shutdown margins and/or pcwer i level and, the consequences of a stuck rod, with only 2 RC loops operating instead of four. It is conceivaole that two loop operation may be less conservative than either 4 RCPs continuing to operate or 4 RCPs tripped on Safety Injection, due to an increased cooldown in the core due to circulation l I

(compared to the tripped case) but a much decreased core flow rate to handle the event. The potential short term consequences of bulk voiding and loss of ,

circulation in the non-operable loops cannot be ignored. .

If during cooldown, an MSLS cools the RCS down to 212*F e.g., the residual shutcown will be at 1% delta k/k whereas the proposed T.S. margin at Zero Power according to T.S. Page 3/4 1-1 was 1.6 delta k/k. Please clarify, and i et wnat condition during coolcown the 1.6% delta k/.k is reacned.

I Given the circumstances that the " Operating Instructions" described above are not a part of the proposed T.S., any T.S. allowing operability of less than 4 RCS Loops in MODE 3 would be in non-confonnance with the current Licensing Basis Safety Analysis in the FSAR in a non-conservative manner, and the licensee would be required to evaluate and propose.

For this licensing basis event, from Zero Power, Reactor Trip does not occur on Power Flux Trip, but on' Pressurizer Pressure-Low (SI) (acove P-11) [ reference our required confirmation of this in an earlier item] so the Power Flux Trip .

is not recuired to be Operaole. ._

At less than P-11, these circumstances are changed for the MSLS, and Reactor Trip does not occur until Containment-Hi is achieved, for a break inside con-tainment.

For a break outside containment, however, high negative steam rate isolates main steam isolation valves only, but their is no Safety Injection, no Reactor Trip (on SI), and under the exisiting proposed T.S. no safety relatec Reactor Trio System Instrumentation of any nature to Trip the Reactor and Insert tne movable control rods to benefit from potentially increased availacle snutcown-margin. In addition to all this, the licensee proposts that MSIV closure times under these conditions in Not Applicaole. -

Given the circumstances of the prooosed T.S. , and T.S. allowing CPERABILIT/ of less than 4 RCS Loop in MODE 3 under these circumstances would be in noncon-

{ formance with the current Licensing Basis FSAR in a nonconservative manner, and the licensee would be required to evaluata and propose.

Additional events which exhibit a rapid cooldown and depressurization of the RCS; are:

a) Accidental Depressuri:ation of the main steam system at no load, (reference 7, page 15.2-35, revision 36).

b) Minor Secondary System Pipe Breaks [at no load]; reference 7, page 15.3-4 revision 27). ,

63 Revision A 06/01/B4

r g(c G.2.6.3 LOSS OF PRIMARY CCOLANT: OCCURRENCES 1

Concerning: "Small Break LOCA" This is ciscussed in reference 7, section 15.3.1 for a S3LCCA frca ratac power, and reference 8, item 212.75 page Q 212-47b for a SBLOCA cetween RCS concitions of 1900 psig and 1000 psig/425"F in Hot Standby, and Q 212-64, item 3 together I with SER Supp. No.2, reference 12, page 6-8 for the remaining situations. See also in general, reference 12 pages 6-6 to 6-8 in respect of ECCS System

, Performance Evaluation from Hot Standbye to and including RHR.

The FSAR analysis for SBLOCA in reference 7, Section 15.3.1 states that:

"During the earlier part of the small break transient, the .

effect of the break flow is not strong enough to overcome

,the flow maintained by the reactor coolant pumos .through the core as they 'are coasting down following tr'ip: there-fore upwarc flow througn the core is maintained."

Tooical Recort, WCAP 8356 (reference 19) is the basis (reference 8, page Q 212-47b last paragraph) for the SBLOCA calculations to the same reference 8.

These were undertaken with all pumps initially running followed by either i

a) all pumos tripped or b) continuing to run. The general conclusion from this report, reference 27, page 4-31, is that: ,

"Due to the action of the running (non-tripped) pumps, less negative core flow occurs frca the flow reversal compared to -

the case [ ] where pumps are immediaely tripped." and "The net result of these effects is a smaller peak clad temper-ature for the pumps running case compared to the pumps ~

tripped case. Hence, for ECCS analysis for W 4 loop plants the reactor coolant pumos are assumed to be tripped at the initiali:ation of a postulated LOCA and a locked rotor pump resistance is used for reflood."

At this time therefore, tne NRC must conclude that RCS pump coeration and coast cown is i=::ortant to reducing the loss of core level subsecuent to the event; also in maintaining unsecaratec two pnasa flow conditions anc in ensuing racid Boron (mixing and) Injection to the core. Rapid baron injection .ould not be an important issue if boron concentrations are already at cold snut down values, .-

but minimi:ing loss of core level is important.

Until further evaluations are made, we must conclude that the current Safety Analysis Limits of the SBLOCA event is 4 RCS pumos OPERABLE in MODE 3 down to 425 psig/250*F. The current proposed T.S. are therefore non-conservative and the licensee must evaluate and propose.

\ Given the circumstances of the proposed T.S., operability of less than 4 RCS Loops in MCDE 3 would be in non-conformance with the Current Safety Analyses Limits in a non-conservative manner and the licensee is required to evaluate and propose.

l 06/01/S4 64 Revision A

T C. ( Cs!

Additional events of a similar nature to the SBLOCA events include: e a) Accidental Depressurization of the Reactor Coolant System (reference 7, l page 15.2-33, revision 7).  ;

b) Steam Generator Tube Rupture (reference, page 15.4 - 13a, revision 38).

c) Rupture of a Control Rod Drive Mechanism Housing at Zero Power (reference 7, i page 15.4.6, revision 42).

Both events, a) and b), are analyzed in the Licensing Bases at Full Power, and .

use Pressurizer Pressure-Low as a first reactor trip. At zero power, with current proposed T.S. this reactor trip is proposed as Not Operable.

~

For event c), from Zero Power, Power Range Neutron Flux, High Set Point Trips the Reactor;. Pressurizer Pressure-Low (SI) initiates Safety Injection; . .

reference 7. page 15.4-29,* revision 43,. paras. I and 5. Whereas both these

{

protections are proposed by the T.S. in MODE 2, they are not proposed for MCDE 3 which differs from the circumstances of MODE 2 by only a marginal reduction in RCS Temperature.

< The FSAR, reference 7, Table 15.4.6-1, revision 42, shows this occurrence as being the only event at Zero Power, analyzed to a smaller N' of RCPs than 4; it has been analyzed for 2 only. This is an accident with substan-tial but " acceptable to Condition IV occurrences" consequences in terms of fuel cladding damage and RCS overpressurization, but it required at least two RCPs to achieve that (in the Licensing Basis). Even the two RCPs required --

in this event are not proposed as being required for MODE 3.

The proposed circumstances in MODE'3 are clearly non-conservative with resoect to the Licensing Bases. The licensee shall evaluate and propose.

Concerning the Large Break " Loss of Coolant Accident."

This is discussed in Accident Analyses in Reference 7, section 15.4.1 for a LOCA from rated power; in Reference 8, item 212.75 page Q 212.47, for a LCCA between RC3 conditions of 1900 psig and 1000 psig/425*F in Hot Standbye; in item 212.90(6.3), page 212-61, for a LCCA at and less than 1000 psig/425' in Hot Standbye, and on page Q 212-61b, item 29 for a LOCA in the RHR Mode at ""

425 psig/350*F.

As for the Small Break LOCA, these analyses are presumably based on a RC5 icco coeration, with in general, loss of power to RCS Pumps on Safety Injection.

The large break LOCA analyses used the Topical Report WCAP-8479, reference 7, page 15.4-1. At this time, we expect no difference in the importance of RCPs to that discussed under the paragraph commencing "Concerning Small Break LOCA" which used the W Topical Report WCAP 8356 (reference 19) and which applied to both Large ard Small Break LOCAs.

65 Revision A 06/01/84

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Given the circumstances of the proposed T.S. , any T.S. allowing.0PERABILITY of less than 4 RCS Loop in MCCE 3 would be in nonconformance with the Licensing Basis FSAR in a nonconservative manner, and the ifcensee is required to eval-uate and procese.

G.2.6.4 OCCURRENCES CAUSING AN INITIAL INCREASE OF RCS TEMPERATURE f

Those events causing increases in RCS temperature are of concern because of the potential influence of the positive moderator temperature coefficient resulting from the increased baron concentration. These could be:

a) Main Rupture of a Main Feed Line (Reference 7, page 15.4-10, revision 30),

although tnis is normally evaluated at Rated power with no provision for evaluation as zero power.

b) Start up of an Inactive Reactor Coolant Loop ,

c) Loss of Offsite Power (reference 7, page 15.2-19, revision 7) f d) Partial Loss of Forced Reactor Coolant Flow (Reference 7, page 15.2-16, revision 7) e) Complete Loss of Forced Reae.or Coolant Flow (Reference 7, page 15.3-7, revision 7) -

Except for item b; all these events are licensing bases events from Rated power, and not zero power, so that their importance would normally be minimal except for the po.,1tive Moderator Temperature Coefficient and the complete lack of Safety Related Reactor Trip protection proposed with the Reactor Trip System Instrumentation T.S.

At this time we see no protection against positive temperature coefficients in

f. MODE 3 (4, 5 L 6].

Given the circumstances of the proposed T.S. , Coeracility of less than 4 RCS Loops in MCCE 3 would be in non-conformance with the current Safety Analyses Limits in a non-conservative manner and the licensee is required to evaluata

_and propose.

~

Ye G.3 CONCLUSIONS Occurrence II, III and IV Events in MODES 3, 4 and 5, can result in returns to power with hign peaking coefficients requiring effective reactivity control and/or reactor core flow for RCS protection, including DN8R, at tne very substantially reduced pressure levels in the loop (2250 psig to 425 psig and less]. Concomitant decreases in RCS temperatures are beneficial, but the ir 3rtance of RCS pressure may be dominant. Acceptable RCS protection there-fore requires RCS flows which are substantial, and/or effective reactivity f control including combined action to limit potential reactivity excursions.

At this time, with the prooosed T.S., 4 RCS loops (with increased Reactor Trio Protection) would be recuired at entry into and during MODE 3 to meet the requirements of just the Licensing Basis Events From Zero Power. In MCCE 4, 06/01/84 66 Revision A

ye wr e) operation of 4 RCS Loops, whilst on RHR, may be undesirable because of the substancial additional burcen on the RHR system; so, nonoperability of all ,

RCPs must be compensated by other controllable factors such as inserting all movable control assemblies and removing power from the Reactor Trip System Breakers, closure of Main Feedwater [ Containment] Isolation valves to both Main and Auxiliary Feedwater Systems, Closure of Main Steam Isolation Valves,  !

and Boration Control measures additional to those included in the proposed T.S.

i

}

An additional available alternate action is to use, within MODE 4, a minimum set of RCS pumps (and loops) as established by Safety Analysis, to cool the a plant down to effectively zero pressure (gauge) in the Steam Generators [or less if the condenser was still available] t>efore transferring the heat sink -

to the RHR system. This would ensure control of Steam Line Break, and LOCA events, small and large, down to RCS conditions where RCS flows are not r necessary.

The current T.S. are nonconservative in respect to the Licensing Basis in '

t respect to these concerns. .The Licensee snall evaluate and' propose. J 1 k

T.5. SECTION 3/4.4.1: RCS LOOPS AND COOLANT CIRCULATION START UP ('dODE 2) AND POWER OPERATION (MODE 1).

The LCO requires all [4] reactnr coolant loops to be in operation in MODES I & 2.

The ACTICN S'tatement requires that in the event of loss 'of 1 (of 4] RCS Loop in MODES 1 & 2, the licensee is required to be in at least HOT STANCBY within 1 hr.

The current Safety Analysis Limits in the FSAR, reference 7 page 15.2-16, revision 7, requires an immediate trip of the reactor to RTI & ESFAS response times in the event of loss of 1 RCS pump. Also, placement of the RCS in Hot Standby with less than one loop operable [without other ccmcensating condi-tions) would be non-conservative in resoect of the existing FSAR.

The Action Statement is non-conservative with respect to the current licensing basis and the licensee shall evaluate and propose.

T.S. surveillance requires verification of Reactor Coolant Loop (RCL) circula-tion once every 12 nours. This is unacceptable consicering the Safety Analysis -

limits required above for loss at one pump. In the event of failure of the Low Reactor Coolant Flow Reactor Trip; the operator should respond immediately to the related Alarm to trip the reactor, if it remains. Reference to earlier work of this review will show that there is no alternate, or diverse, sensor for low flow in one Reactor Coolant Loop. Further the FSAR analysis does not provide an evaluation of the consecuences of a 10 min delay by the operator on hearing the Alarm - if it has remained operable from available [3 channel]

LCGIC. Additionally, the FSAR proposes no alternate trips for the reactor, with related svaluation, such as over temperature leading to Pressurizer Level-High and Pressuri:er Pressure-Hign. The Action Statement would place the pla6t outside the current licensing basis for normal operation and is non-conservative with resoect to that. The licensee shall evaluate and propose.

67 Revision A 06/01/84 k

i . .

Further it can be proposed, for this event analyzed in ref. 7, page 15.2-16, revision 7, that Criterion 22, Protection System Independence has not been a,et:

" Criterion 22- Protection system independence. The protaction system shall be designed to assure that the effects of natural pnenomena, and of normal operating, maintenance, testing, and postulated accident conditions on redundant channels do not result in loss of the protaction function, or shall be demonstrated to be acceptable on some other defirtd basis.

' Design techniques, such as functional diversity or diversity in component design and principles of operation, shall be used to the extent practical to prevent loss of the protection function." -

TheFacilityisnon-conservativewithrespecttothisbegulation,thelicensee '

shall evaluate and propose. This is a generic issue.

The surv'illance e requirement, every ,12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />, is intendea to ensure not.only that the system is operating, but that it is operating at process conditions wnich can be evaluated to show that the equipment is capable of performing its Licensing Basis Safety functions. The proposed T.S. requirements are absent in this inforsation; it is therefore non-conservative and the licensee shall evaluate and propose.

T.S. P:ce 3/4 4-2: RCS HOT STAN08Y .

The current T.S. requires only 2 RCS loops to be in operation in this MODE 3.

The basis for this requirement on TS Page S 3/4 4-1 says only: "In MODE 3, a single reactor coolant loop provides sufficient heat removal capability for removing decay heat; however single failure considerations require that at least two loops be CPERA8LE." This basis is unacceptable since the facility

' is required, within this condition of normal operation, and its existing licensing basis, to also be able to withstand related valid Condition II, III and IV occurrences; and earlier work has shown the Safety Analysis Limits for the plant currently requiring at least 4 RCS pumps for this MODE.

The Action Statement allcwing 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> with only one RCS loop operable is non-conservative with respect to the current Safety Analysis Limits.

At this time, any No. of loops less than 4 in M00E 3 is non-conservative with respect to the existing FSAR and the plant should be transferred to operation ,.

in MCOE 4 under these circumstances, with approved maximum normal coolcown rates.

It is recognized there are many protective actions which may provide more flexibility in this MODE within NRC/RCS Safety Criteria but they are not included within the current T.S. proposed by the licensee; further that final ch .ca of such actions may be determined by " additional" protective procedures already in placa at the plant, but not included in the T.S. where they are required by 10 CFR 50-36. Also, the particular comoinations of protections which could be proposed may depend on providing the facility with maximum flexibility in other operations in this MODE 3 consistent with meeting Regula-tory Safety requirement. See our earlier review under General.

i i

06/01/84 . 68 Revision A

Given the circumstances of the proposed T.S., operability of less than 4 RCS foops in MCOE 3, HOT STANO3Y, would be in non-conformance with the current

• Safety Analysis Limits in a non-conservative manner and the licensee is required to evaluate and procose.

It further follows, that the proposed surveillance requirement T. S. item l 4.4.1.2.3 that at least one reactor coolant loop shall be verified in operation and circulating reactor coolant at least once 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> is also invalid and 5 should be changed.

The surveillance requirement, once every 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />, is intended to ensure not -

only that the system is operating, but that is is operating at process condi- ~

tions wnich can be evaluated to show that the souipment fs capaole of performing its Licensing Basis Safety Functions. The proposed T.S. requirements are aosent in this information; it is therefore non-conservative and the licensee shall .

evaluate and propose. ,

• I Surveillance recuirements for the S.G. call for a level'of 12*. at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. This is not in accordance with the Licensing Basis; this level j

is the S.G. Low - Low Trip Set Point. All conditions II, III and IV occurrences g require in general, for this S.G. level to be at the programmed Set Point for the Zero Power Condition with automatic actuation; we have no evaluation at alternate conditions. Therefore this exlisting proposal is outside the current Licensing Basis and non-conservative. Reference our earlier comments uncer Item 2.1.1, Item f. The licensee shall evaluate and propose.

"This Footnote proposes that; in HOT STANOBY (MCDE 3):

"*All reactor coolant pumps may _be de-energized for up to 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> provided:

(1) no operations are permitted that would cause dilution of the Reactor Coolant System boron concentration, and (2) core outlet temperature is main-tained at least 10*F below saturation temperature."

This is a natural circulation condition; the only Licensing Basis calculation for this is the Natural Circulation calculations of reference 7, page 15.0-27

" Lass of Offsite Pcwer to Station Auxiliaries"; but at McCE 2 Zero Power ::nci-

' tions with related programmed process conditions of Zero Load Pressure and Temperature in the locos. No basis is provided for ensuring that natural circulation will be safe over the range of conditions now expected in this MODE 3. Earlier considerations show that more comorehensive protections "

against the possibility of Condition II. III and IV occurrences must involve.

in addition to isolation of all boron dilution sources, securing Reactor Trip System Breakers in the Open Position, closure of MFW isolation valves, iscia-tion of MSIVs and possibly an optimum baron concentration. At present, the only Licensing Basis for controlling this particular situation is the Emergency Operating Guidelices.

Given the circumstances of the proposed T.S., the proposal to de-energize 4 RCPs for up to one hour is outside the Safety Analysis Limits of the FSAR and is non-conservative with respect to that.

~he licensee shall provide the reason for this requirement including the

• expected condition of the Facility, and'then analy:e, evaluate and propose.

69

levision A C6/01/84

l Earlier concerns under General 2.6.1 addressed the need to evaluate the con-seguences of the Start Up of an Inactive Reactor Coolant Loop in this MCDE. No apparent T.S. provision has been provided in the prooosed T.S. The licensee fshallevaluateandpropose.

Action item b. states:

"b. With no reactor coolant loop in operation, suspend all operations involving a reduction in boron concentration of the Reactor Coolant

. System and immediately initiate corrective ACTICN to return the required reactor coolant loop to operation."

{

This instruction is invalid. The only Licensing Basis action available is the Emergency Operating Guidelines for the Natural Circulation. This procosal .

is non-conservative with respect to the Licensing Basis. The licensee shall t evaluate and propose. -

_2 T.S. Pace 3/4'4-3. REACTCR COOLANT SYSTEM - HOT SHUTCO$N.

The proposed T.S. should be supplemented by the conditions contained within the brackets [ 3:

"3.4.1.3 At least two of the reactor coolant and/or residual heat removal (RHR) Icops listed below shall be OPERABLE [and energized from separate power divisions] and at least one of the above reactor coolant and/or RHR locos shall be in operation:*" [ Additionally two RCS loops must always be OPERABLE whenever RHR loops are in operation]

~

a. Reactor Coolant Loop A and its associated steam generator [ including related auxiliary faedwater pumps] and reactor coolant pump,*

b. Reactor Coolant Loop 9 and its associated steam generator [inclucing related auxiliary feecwater pumos] and reactor coolant pumo."

c. Reactor Coolant Loco C and its associated steam generator, [ including relating auxiliary feecwater pumps] and reactor coolant pumo,"

d. Reactor Coolant Loop D and its associated steam generator, [inclucing related auxiliary feeawater pumps) and reactor coolant pumo,"

e. RHR Loco A,"*" and

f. RHR Loop B."""

/0 APDLICAB b b: MODE 4. [Less than 425 psig/350*F]"

The licensee shall evaluate as outlined earlier under Item, General, for RCS locos operability requirements and make proposals relative to the status of many elements of the protection and operations system to ensure that RCS safety is maintained for related Condition II, III and IV occurrences. At this time, witn tne procosed TS in which limited boration is used and Reactor Tric System afety Related Instrumentation and Safety Injection Instrumentation are all but 06/01/84 70 Revision A

.. .. , _ -_ _ - _ - . _ _ . - - - . _ ~ --

t

/0(c}'-))

eliminated, the safety status of the facility is outside the Licensing Basis .

of the FSAR in a non-conservative manner.

Each of the OPERABLE loops, whether RCS or RHR, are to be enorgized from

' separate power divisions to protect against single failure of a bus or distri- l' bution system. When the RCS systems are useo, the related Auxiliary Feedwater ,

systems are also required to be operable. j 3

! The additional requirement proposed, for two RCS loops to be operable whenever i RHR loop /s are in operation, is based upon reference 8, page Q 212-55 and 56, to provide for the failure of a single motorized valve in the RHR/RCS suction  :

line in both MODES 4 and 5 and possible non-availability of offsite power sources. The FSAR provides, that on failure of the valve:

I. "Approximately 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> are available to the operator to establish an alternate means of core cooling. This is the time it would take to heat I the availaole RCS ' volume from 350*F to the saturation temperature for .

1 400 psi (445*F), assuming the maximum 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> decay heat load.

Torestorecorecoolinh,theoperatoronlyhastoreturntoheatremoval I via the steam generators. The operator can employ either steam dumo to the main condenser or to the atmosphere, with nakeup to the steam genera-tors from the auxiliary feedwater system. The time required to establish the alternate means of heat removal is only the few minutes necessary to , , ,

l open the steam dump valves and to start up the auxiliary feedwater system." {

The APPLICA8ILITY MODE 4, is necessarily qualified by [less than 425 psig/350*F] --

by the LOCA analyses already referenced above under our review Section 3/4 4.1 Subsection G.2.6.3 "Concerning Large Break Loss of Coolant Accident." See reference 8, page Q 212-47.d wnere it is described that "After several hours into the cooldown procedure (a minimum time is )

I accroximately four hours) wnen the RCS pressure and temcersture have J cecreased to 400 psig and 350*F."

i And arising from a later revision 25, the FSAR advises on page Q 212-315 revi-

sion 29 concerning ECCS calculations in a later submittal uncer Revision 2*

that

~'

"The response provided in Revision 28 addressed the subject of operator

) actions and ECCS availability. Consistent with the information proviced in Revision 28, a postulated LOCA in the RHR mode at 425 psig RCS pressure has been assessed." J

! The additional Action statement that:

b. "With no reactor coolant or RHR loco in coeration, suspend all operatient involving a reduction in boren concentration of the Reactor Coolant System and immediately initiate corrective ACTION to return the recuired coolant loop to coeration."

i i

06/01/84 71 Revision A

._ , _ .____ ._, _ . - _ ___ .__._ _____ M_L __,_ : _ _ -~. . . . _ . _ . . _ . _ _ _ _ . _ _ , _ - . _ -

I I

t 5

and the additional notation that

"**"All , reactor coolant pumps and RHR pumps may be de-energized for up to I hour provided: (1) no operations are permitted that would cause dilution of the Reactor Ccolant System baron concentration, and (2) core outlet temoera-ture is maintained at least 10*F below saturation temperature."

are unsupportable by present analyses in the FSAR. These proposed T.S.s are the same as for MODE 3 and our relevant comments and requirements under T.S.

Page 3/4 4-2: RCS HOT STAN08Y should be applied to MODE 4. Emergency Oper-ating Guidelines Apply. This proposed T.S. is non-conservative with respect to the Licensing Basis. The licensee shall provide the reason for the require-ment including the expected condition of the facility, and then analyze evaluate an p

"-^g,qoppse.

l O (f Surveillance requirement 4.4.1.3.2.should verify S.G. water level at the Safety ~')

, Analysis Limit for the Licensing Basi.s. which is the no-load programmed level, .

not the current proposed TS value which is the S.G. Low-Low Level [ Reactor Trip] and AFW actuation. This proposed TS is non-conservative with respect

! to the current Safety Analysts Limits and the licensee shall evaluate and propose.

Surveillance requirement 4.4.1.3.3 verifying one loop in operation every 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />, is unsupportable as all protective trips on low flow in the RCP. loops in this condition have been removed. If low flow channel trips on the RCP loops are not required to be operable why should the related Alarm be operable. A low flow alarm for the RHR has been provided by the FSAR under reference 8, page Q 212-56, ites: .

" Case 1: The Reactor Coolant System is closed and pressurized.

! The operator would be alerted to the loss of RHR flow by the RHR low flow alarm. (This alarm has been incorporated into the McGuire design).

Since currently, these two types of alarms are the only means of alerting tne operator to a Loss of Flow condition in the loco, which is beyond tne Safety i Analysis Limits, then the alarms on botn the RCS and Loop Flows snoulc ce Safety Related and included within the T.S.; and without further analysis at l this time, two loops should be placed in oceration. A proposal is made by the

! NRC for low flow alarms in each of the secarated cooling systems, under Proposed -

T.S. Page 3/4 4-6a of this review. Regular surveillance should ce proposed to ensure they remain operable as appropriate, over a specified surveillance period.

l The Surveillance requirement, every 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> is intended to ensure not only i i that the system is operating, but that it is operating at process conditions i which can be evaluated to show that the equipment is capable of performing its

. sign basis Safety Function. The current surveillance requirements for this item, i.e., for the RCS and RHR systems in Hot Shutdown in T.S. Item 4.4.1.3.3, are absent this information; it is therefore non-conservative and the licensee shall evaluate and propose.

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06/01/84 72 Revision A i

9

. e lO ( w_^'

I Item 4.4.1.4.4 (Proposad). It is proposed that an additional item be inserted j which reads: "The related auxiliary Feecwater System shall be determined CPERABLE as per the recuirements of T.S. 3.7.1.2 (and 3.7.1.2.a as acclicable)."

Current proposed T.S.s on T.S. page 3/4 7-4 are non-conservative in this mattar

! by not providing any operability requirements for AFW in this MCDE. The licensee shall evaluate and prepose.

An additional item is also required in which Atmospheric Dump Valves operability -

is established. The current T.S. are non-conservative in this matter; they make no provision for operability of this item (see later proposed T.S. page 3/4 7-8a). [ General comment: Operability of each of S.G. water level, AFW and ATMOSPHERIC OUMP VALVES in this MODE is probably better defined under each of these items in their particular sections of the T.S. See later sections of this review as identified above.]

The FSAR addresses the tensequence of a failure, closed of the ,1' solation valve in the RCS/RHR line; it accresses the analysis from 350*F in the RHR MODE wnen a bubble is present in the pressurizar. This will also be valid'down to the RCS temperature at which the bubble will be established, i.e., below 300*F according to reference 19, page 52-21a, revision 33, first para. If the licensee does operate the plant so that the system is water solid between 200*F and 300*F fn MODE 4, a loss of conting could result in a potential overpres-suri:ation of the system and the reviewer is not aware of a'ny evaluation of the adequacy of the existing Low Temperature Overpressure Prot'ection System to accommodata that event. The licensee shall evaluate and propose.

T. S. Dace 3/4 4-5: COLD SHUTDOWN rMCOE 5] WITH LOOPS FILLED.

The current preposed T.S. provides:

3.4.1.4.1 At least one residual heat removal (RHR) loop shall be OPERABLE and in oceration", and either:

a. One acditional RHR loop shall be OPERA 8LE#, or

b. The secondary side water level of at least two steam generators shall be greater than 12P..

The current FSAR requires two (2) OPERABLE RHR trains on two (2) recuncant '"

electrical buses so that each pump receives power from a different source, reference 20, Pages 5.5-24 In the event of Loss of Offsite Power, the pumps are automatically transferred to a separate emergency diesel power supply.

Therefore; the current licensing basis is that 2 residual heat removal loops shall be operable. The above provision for either an RHR loop or two steam generators is therefore not in accordance with the Licensing Basis. The proposed T.S. in this respect is also non-conservative as it would necess- ily require S.G. temperatures greater than 212*F (Atmos Press in SGs) wnica would place it outside the Cold Shutdown MODE into the Hot Shutcown MODE - which is outside the recuired Functional MODE.

The T.S. requirement for one RHR loop in coeration and one to be available OPERABLE is currently not sucportable by analysis evaluating the situation in whicn all RHR cooling is lost in a water solid condition; reference our 06/01/84 73 Revision A

immediately proceeding item T.S Pace 3/4 4-3. In this case, if one only RHR loop is operating, loss of that single loop cause overneating in a water solidstate with potential overpressurization. Does the alarm of loss of RHR Flow which is required, and an coerater resconse time of 10 mins, provide sufficient time to commence operations of the second RHR loop to the extent necessary to mitigate the consequences of any potential overpressure event in an acceptable manner. The licensee shall evaluate and propose.

Use of secondary side water level of at least two steam generators is discussed in reference 14 for circumstances in which the RHR is isolated from the RCS and its final acceptability for licensing purposes is still not resolved. '

This, in addition to its temperature limitation means that it cannot be proposed

• as an alternate means of remov.ing decay heat during Cold 3hutdown. The proposed T.S. is therefore not in accoraance with current Safety Analysis Limits, and also non-conservative.

.As discussed in the pre'vious item T.S. Page 3/4 4-3, what is required by the current Licensing Basis in Mode 5, is to have availacle two OPERA 8LE RCS loops (including AFW, SG and SG/PORVs] to meet the circumstances of failure closed of the RHR isolation valve and in which case the RCS returns to MODE 4 with its particular MODE 4 requirements as discussed earlier. The absence of this as an LCO requirement in the proposed T.S. makes it non-conservative with respect to the Licensing Basis. The Licensee shall evaluate and propose.

~ .

Footnote *: This item proposes that an only avai1able operational RHR pumo may be de-energized for up to 1 hr. This event has not been evaluated, is not within the Licensing Basis, and is non-conservative. The licensee should define the circumstances, analyze and evaluate and propose.

The propcsed'survef1' lance requirement /4.4.1.4.1.2 provides that "At least one RHR loop shall be determined to be in operation and circulating reactor coolant at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. The items of significance here are Operable Safety Related Flow Alarms with a surveillance frequency ensuring high procability of alarm in the event of an RHR flow failure, and a related concern for overpres-sure protection and recovery. The licensee shall evaluate and propose.

The surveillance requirement, every 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />, is intended to ensure not only tnat the system is operating, but that it is operating at process conditions wnich can be evaluated to show that the equipment is capacia of performing its Licensing Basis Safety Function. The current requirements for this information '

for the'RHR systems in T.S. 4.4.1.4.1.2 are absent; it is therefore non- '

conservative with respect to the Licensing Basis. The licensee shall evaluate and propose.

T.S. Pace 3/4 4-6. REACTOR COOLANT SYSTEM - COLD SHUTDCWN. LOOPS ARE NOT FILLED Item 3.4.1.4.2 requires that:

"3.4.1.4.2 Two residual heat. removal (RHR) loops shall be OPERABLE # and at least one RHR loop shall be in operation.*"

Additionally, the current FSAR requires that each of the RHR trains be proviced with power from (2) redundant ele.ctrical buses so that each pump receives l 06/01/84 , 74 Revision A

s t

power from a different source; reference 20, pages 5.5-24, revision 9. Without this recuirement, the T.S. is less conservative than the FSAR and the licensee

• shall evaluate and propose. .

l Additionally, the current FSAR, reference 8, page Q 212-57, revision 25, describes that in the event of loss of flow caused by isolation of the RHR/RCS Isolation valve (and also by cessation of flow in the system]

"The operator would be alerted to the loss of RHR flow by the RHR low g flow alarm. ,

Assuming worst casa conditons (maximum 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> decay heat, air in the .

steam generator tubes, and the RCS drained to just below the vessel i flange) and making conservative assumptions about the amount of water availacle to heat up and boil off, if the operator took no action, boiling '

woulo begin in about five minutes, the water level in the vessel would be down to the level of fuel in about.100 minutes, and the pressure would ..

increase to 550 psi in about 40 minutes (the pressure

• rise could be timited to about 550 psi by opening the pressurizer power operated relief valves)."

In the event only 1 RHR loop is required to be in operation,the LCO should therefore reouire 2 operable Safety Related RHR flow alarms on each single operating RHR system so that the operator can respond within 10 mins to com-mence operation of the redundant system. However, this time frame is exces-sive sinc. boiling will have commenced. It is necessary to maintain two operating RHR systems so that boiling may be eliminated on single failure. . .

The licansee shall evaluate and propose.

Additionally, the above information defines an LCO of a minimum volume of water for thu related event in which the RCS is drained to just below the Reactor Vessel flanges and which minimum volume shall be included in the T.S. as an LCO with appropriate surveillance and Action Statements. A further T.S. require-ment is tnat any such min volume should be such that the level of water in or above the RCS loops be such as to provice acceptable flow, inclucing NPSH conditions, over the range of temperstures expected, at inlet to the RHR pumps.

Absent those recuired conditions from the Limiting Conditions of coeration makes them non-conservative in respect to the Licensing Sasis. The licensee shall evaluate and propose.

~

Concerning Action item b., this provides that'

b. With no RHR loop in operation, suspend all operations ~ involving a reduction in baron concentration of the Reactor Coolant System and immeciately initiate corrective ACTION to return the required RHR loop to operation.

Further: In the event that RHR cooling cannot be restored in " sufficient" time, the FSAR states that, in the event of loss of flow caused by the single RCS/RHR motorized valve:

"To restore core cooling, the operstar would first attempt to fill anc pressurize the esactor coolant system with the centrifugal charging pumps. If the system can be pressurized to the range of 400-500 psi, the 06/01/84 75 Revision A

operator could return the plant to heat removal via the steam generators.

To do this the operator would have to jog the reactor coolant pumes to sweep the trapped air from the steam generators. He would also have to ocen the steam duma valves (to atmosenere or the main con::enser) anc start up the auxiliary faecwatar systam."

In this M00E therefore, it is necessary to ensure that 2 RCS 1 cops with operacle SG, AFW supply and SG/PORVs are operable from separate buses, to be available, in the event of the single failure discussed. This would also support the general concern in the event of noncapacility of restoring failed RHR systems to Operability within an acceptable time frame, including the possibility of core uncovery in 100 mins. [The licenses shall also reference any Emergency Coerating Guidelines in this respect). Without provision for RC3 Loop Coera-oility required by the Licensing Basis FSAR, the current T.S. LCOs must be -

considered non-conservative with respect to the Licensing Basis, and the licensee shall evaluate anc propose. ,

Item 4.4.1.4.2, A surveiliance requirement, specifies:

At least one RHR loop shall be determined to be in operation anc circulating reactor coolant at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

A time delay of 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> is excessive to verify a loop in operation, and this has been considered earlier in this section. Further the surveillance require-ment, every 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />, is intended to ensure not only that the system is operating, but that it is operating at process conditions, including instrumentation and control, wnich can be evaluated to show that the equipment is capable of performing its design basis Safety Function. The current requirements for this T.S. Item are absent in this information; it is therefore non-conservative and the licensee shall evaluate and propose.

Footnote *: Provides that,

""The RHR pumo may be de-energi:ed for uo to I hour proviced: (1) no ocers-tions are permitted that would cause dilution of tne Reactor Coolant System boron concentration, and (2) core outlet temoerature is maintainec at least 10*F below saturation temoerature."

This departure from the Licensing Basis of two availaole RHRs witn effective _

cooling at all times it outside the FSAR Licensing Basis in a non-conservative manner. Further this is also supported by the earlier information of this section that boiling would commence in 5 minutas with core uncovery in 100 minutes. The provision is outside the Licensing Basis in a non-conservative manner and the licensee shall evaluate and propose.

T/S Dice 3/4 4-6(a) Prooosed.

A new subsection should be added entitled " REACTOR COOLANT SYSTEM, HOT SHUTCCWN TO REFUELING, APPLICABLE MODES 4, 5, & 6 which recuires a LIMITING CON 0! TION OF OPERATION tnat two RHR Flow Alarms to Safety Related requirements shall be operable on each RhR loop when only one RHR loop is in oceration under the provisions of the Technical Specifications. Appropriate Action Statements and surveillance requirements shall be applied.

06/01/84 76 Revision A

The safety basis for this was established in the FSAR, as indicated in earlier

• sections, and the need for safety related redundancy arises to ensure RCS integrity to Safety Related Criteria as discussed aoove. The current T.S. is non-conservative with respect to tne Licensing Sasis.

T.S. SFCTION 3/4.4.2 SAFE / VALVES SHUT 00WN (MCDES 4 and 5) ,

The T.S. requires that:

"3.4.2.1 A minimum of one pressurizer Code safety valve shall be OPERABLE I with a lift setting of 2485 psig 1*.. a  :

I Apo' !CABILITY: MCOES 4 and 5.

AC'TICN: f With no pressurizer Code safety valve OPERA 8LE, immediately suspend all operations involving positive reactivity changes and place an OPERA 8LE RHR  ;

loop into operation in the shutdown cooling MCDE."

Reference our review comments and requirements under T.S. 3/4.4.2 SAFETY '.

VALVES, OPERATING which are also applicable to this section. The current T.S.

must be considered nonconservative with respect to the Licensing Basis. The Licensee shall evaluate and proposa. .

The Action statement is based (reference T.S. page B 3/4.4-2) on the premise that INOPERA8ILITY of the Safety Valve in Modes 4 and 5 needs to be offset my operacility of pressure relief vaIves in the RHR systems. This is not tne safety basis for Action. The safety basis is, that the Reactor Coolant Pres-sure Soundary has been effectively rendered inoperable requiring the operator to croceed to a cold shutdown condition with :he zero pressure (gauge) in corn RCS and SG systems, and related reactivity control actions to ensure inat no return to nuclear power is possible. This needs to ce done in a manner consistent with the nature of inocerability of the Safety Valve. The current T.S. is nonconservative with respect to the Licensing Basis; tne licensee snait evaluate and propose.

~

Furtner, McGuire Units 1 and 2 do not use RHR overpressure protection of the RCS as the plant utilizes two available PORVs on the pressurizer, reset to 400 psig (reference review under T.S. Page 3/4 4-36) in the primary coolant system. In this respect, the proposed action statement is non-conservative and contrary to the Licensing Sasis. The licensee shall evaluata and prepose.

The Surveillance Recuirements should contain the minimum discnarge capacity required of this valve as defined in the Licensing Basis. They should also ensure the maintenance of satisfactory environmental conditions consistent with reliable valve cperability. The licensee shall evaluate and propose.

06/01/Sa 77 Revision A

Fr .a - --- - -

e .

t e

T.S. Section 3/4 a.2 SAFETY VALVES OPERATING The proposed T.S. requires all (3) pressurizer Code Safety Valves to be Operacle in Appitcable Modes 1, 2 and 3.

The Action Statement requires that

" ACTION:

With one pressurizer Code Safety Valve inoperable, either restore the inoperacle valve to OPERA 8LE status within 15 minutes or be in at least HOT STAND 8Y within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> ano in at least NOT SHUTD0kN within the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />."

.' Failure of the Pressurizer Code Safety Yatv'e, in general, would infrin'ge the -

{ntegrity of the Reactor Coolant Pressure Soundary and the RC3 should be arougnt to the cold shutdown condition, as rapidly as possible, with zero (gauge) pres-sure in both the RCS and SG, in a manner consistent with the nature of the inoperacility, and potential for all positive reactivity levels eliminated.

The worst situation would be 'that of an " Accidental Depressurization of the Reactor " Coolant System" analyzed for the most severe conditions hicluding saximum core power, reference 7, page 15.2-33 revision 7. This type of event would reoutre Emergency Procedures to define the ACTION STATEMENT.

Could other types of failure allow other types of response which could be -

outside the Energency Operating Procedures. The Licensee has not identified others and analyzed and evaluated the related safety to Regulatory Require-monts as a basis for his proposed action.

The T.S. Bases on page 8 3/4.4-2 does not exhibit an accontable understanding

~

of the iscortance of, and potential severity of, the event including failure types and acpropriate Regulatory recuirements including proceoures.

The existing ACTION statement is inaceouate within the Licensing Basis, and therefore unacceptacle. The only existing Licensing 8 asis must so within the analyses recorted in reference 7, page 15.2-33, revision 7, and the preocsed Action Statement does not recognize these circumstances. The existing Action -*"

Statement is therefore nonconservative with respect to the Licensing Basis; the licensee shall evaluata and propose.

LC3 and surveillance procedures must also address position indication and/or di's charge flow seasurement procedures, including pressurizer relief tank condi-tion and other measures to ascertain the operacility of the valve (this is aecessary to satisfy 10 CFR 50 Appendix A, Criterion 20, 32 and 33]. The writer reviewed, in 1983, information pertaining to the GPU/8&W 1awsuit review, and his recollection is that the TMI-2 operatoas " initially thought that the safety valves had developed a leak in the PORys because the valves had lifted on a recent event." There must be a measure of .teceptacle leak tightness fece 06/01/84 78 Revision A se= w

, e measurable parameters "in operation" to ascertain the status of the valve so -

that acceptable measures can be taken.  !

\

The safety basis for the concarn rests not only in the previous position addressed acove, but also, that in the event of failure of control grace " pres-sure control devices" these valves will be challenged on the following occur-rences within the Licensing 8 asis.

I

• Startup of the Inactive Coelant Loop; reference 7 Figure 15.2.6-1,

• revision 4 -

• Loss of Load Accident; reference 7. Figure 15.2.7-5, revision 38

- Loss of Normal Feedwater; reference 7, page 15.2-26 revision 7, para. 3 .

• - Main Feedwater Line areak Accident, reference 7. Figure 15.4.2,.7, ,

revision 34 '

• One Locked Rotor Event; reference 7. Figure 15.4.4-1, revision 32 Safety Valve Operation could also occur on other overpressurization events if same of the early reactor trips fail,to operate as espected. ,

In this matter, the T.S. is nonconservative with respect to Regulatory Require-ments. The Licensee shall evaluate and propose. This could be a generic issue.

~~

Surveillance Requirements should reference the cocuments containing the record of the Inservice Testing of the valves for inspection on a regular basis of 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> so that changing operating staff are kept aware of a potentially changing status on a singularly critical item.

T.S. Section 3/4.4.3 o#tSSLRIZER T.S. Pace 3/4 4-9 The APPLICA8!LITY uc0ES are proposed as 1, 2 sna 3.

Item: Pressurizer Level:

The response of all the analyses of Condition !!, !!! and IV events in refee-ences 7 and 8 depend upon an initial level of water in the Pressurizer =nich is programmed as a varying value dependent upon the Nuclear Power Level. Acci-tionally, tre response of all Condition I events which determine the most conservative set of parameters from which to start Condition !!, !!! and IV events, are also so dependent upon this same programmed pressurizer level.

Since therefore this pressurizer level is used in establishing an acceptacle outcome of these analyses in terms of the issuance of the operating license, they also represent limiting conditions of operation as defined in 10 CFR 30.46.

On this basis therefore, the licensee should provide details of the programmed pressurizer level set points with alloweele values consistent with the relatec channel errers and Safety Analysis Limits used in the FSAR, Section 15 in

, reference 7. The licensee shall evaluate and propose.

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APPLICA8ILITY MODES: Pressurizer level should be graposed for MCOES 1, 2, 3, and 4 (with steam bubble). Down to M00E 4 is provided to cover LOCA and MSLB events considered in reference 8. Also, the plant can then be placed on Automatic Level Control. Aporopriate ACTION and SURvCILU.NCE procedures ,

should be proposac. Licensee shall evaluata and propesa.-

Item: Pressurizer Pressure

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The responses of all the analyses of Condition II, III and IV events in refer-ences 7 and a depend upon an initial value of pressure in the pressurizer (and

'which is not programmed at a varying value in MODES 1 and 2). Additionally, the responses of all Condition I events which detamine the most conservative -

set of parameters from which to start Condition II, III anc'!V avents, are also so dependent upon this same pressurize pressure. ,

Since therefore this.value of pressurizer pressure is used in atta311sning an

• acceptacle outcome of tnese analyses in terms of the issuance of the operating license, they also represent limiting conditions of: operation as defined in 10 CFR 30.44. On this basis, therefore, for each o' MODES 1 through 5, the Ifconsee should provide details of the pressurizer pressure set points with '

alloweele values consistent with the related enannel errors and Safety Analysis Limits used in the Licensing Oasis in the FSAR in Section 15 in reference 7, ard reference S. The licensee shall evaluate and propose.

Accropriate ACTION and SURVEILLANCE procedures should be proposed. The~ licensee shall evaluate and propose.

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T.S. SECTION 3/4.4.4 RELIEFVALVES(POWEROPERATED), i The current T.5 provides that the plant may continue in operation if either one of the combination of 8tock Valve and PCRV is INOPER/48LE. This is a c:ntravention of the regulations which provides uncer 10 CFR 50.2(v) nat:

(v)"Reactorcoolantoressureboundary"meansallthosepressuN-cottaining comoonents of boiling and pressurized water-cooled nuclear power resciars, such as pressure vessels, piping, pumes, and valves wnich are:

(1) Part of the reactor coolant system, or (2) Connected to tne reactor coolant system, up to and including any anc -

all of the following: -

l (1) The outermost containment isolation valve in system piping wnton' penetrates primary reactor containment. '

(ii) The second of two valves normally closed during normal reactor operation in system piping which coes not penetrate primary reactor containment.

(iii) The reactor coolant system safety and reUaf valves.

Since a single failure of either the 81ock valve, or tne PORV. will recuce the level of protection of the Reactor Coolant Pressure Soundary (RCPS) from two 06/01/84 80 Revisien A a

(2) valves to one (1) only valve, the Regulatory Requirements are not met sed '

the plant must proceed to a cold shutdown condition with no potential for positive reactivity changes, within appropriate time frames.

• The current T.S. is nonconservative in respect to Regulatory Requirements.

The licensee snail evaluate and propose.

T.S. Section 3/4 4.5 STEAM GENERATORS i T.S. Pace 3/4 4-11 .

a) 5.G. Levels A nuncer of the Accident Analyses in reference 7 depend ucon an initial level of water in the Steam Generator. A scocific example is the Main Fseewater Line Ruoture Event of Section 15.4.2.2.2 in which AFW auto-start signal on SG low-low level occurs 20 secs are main feedline rupture accurs; reference .

related Taale 15.4-1, page 1 of 4].

Since this, and other events, depend upon a " programmed" water level in the steam generators for an acceptable outcome in terms of the issuance of the operating license, these water levels also represent limiting conditions of operation in respect of 10 CFR 30.46. Please provide details of such SG 1evels including related Safety. Analysis Limits', and respond to the proposition that such values should be included as set Point values and Allowanle values in the proposed T.S. as Limiting Conditions of Coeration for the facility with appropriate Action Statements. The proposed T.S. is nonconservative by their -

aesence, b) Steam Generator Pressures Since Steam Generator Pressures and related Saturation Temperatures under normal staady stata coerstion can be a significant determinant of system responses for Condition II througn IV occurrences analyzed in the Licensing Basis including Section 15 of reference 7, and reference 8, please provice the values used as Safety Analysis Limits in related analyses and again resconc to the proposition that sucn values should be incluced as set Point and Allowamie values as Limiting Conditions of Operation for the facility with accrepriate Action Statements. The proposed T.S. is nonconservative with respect to tne -

Licensing 8 asis, by their absence.

c) Please respond to the proposition that this section should also adecuately icentify the maximum allowable Steam Generator Pressure unoer Transient and Accident conditions with appropriate Action Statements. Maximum SG pressure is one of the Accootance Criteria for safety. The current very limited basis for Steam Generator Pressure integrity is comoletely inadecuate. Please clarify apparent discrepancy between reference 4 Table 5.5.2-1 in which the steam side design pressure for the Steam Generator is given as 1285 psig sno the value quoted in the T;5. Basis Page 8 3/4 7-1 at 1185 psig.

The procosed T.S. is nonconservative with respect to the Licensing Basis, ty

this aosence. .

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4 c) APPLICABILITY MODES 1, 2, 3, and 4:

The current appifcability requirements relate to Structural Integrity considerations.

On inclusion of Steam Generator Level and Pressure as determinants of Osera-bility, the ifcensee should evaluate and propose APPLICABILITY MODES censistent with RCS/SG loep requirements discussed in this review under separate s.ections and particularly under Reactor Coolant Systera and Residual Heat Removal sections in MODES 1 through 5. This will escrace operability requirements from MODES 1, 2, 3 and 4 through 5. The proposed T.S. is nonconservative with respect to i

the Licensing Basis, by the aosence of this information. The licensee shall t

evaluate and propose.

T.S. Pace 3/a 4-36 (REACTOR COOLANT SYSTEM) OVERPRESSURE PROTECTION SYSTEMS T'he cu'rre'nt LCOs require that either of the fol' lowing be Operable;

"(a) 2 PORVs with a lift setting of less than or equal to 400 psig, or (b) The Reactor Coolant system (RCS) depressurized with an RCS vent of greater than, or equal to 4.5 square inches.

The Applicability is MODE 4 wnen the temoerature of any RCS cold leg is less than or equal to 300*F, MODE 5 and MODE 6 with the reactor vessel head on." ,

This section should also include tne often used restraint that:

"A reactor coolant pump shall not be started with one or more of the-Reactor Coolant System cold leg temperatures less than or equal to 300*F unless:

(1) the pressuri:er water volume is less than 1600 cubic feet, or (2) the 4

secondary water temoerature of each steam generator is less than SOF* above each'*of the Reactor Coolant Systam cold leg temperatures.

It is necessary, te exoand the tCOs to all those which should be incorocrated into the opersbility requirements for the pressuri:er and steam generator.cis-l cussed earlier under T.S. Section 3/4.4.3 Pressuri:er and T.S. Section 3/4.4.5 Stean Generators. This additional information defines necessary safety limits for the Licensing Basis event; as in reference 23, wnich is an early Topical

Recort submitted by W for approval. The proposed T.S. is nonconservative in

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the absence of this information. The licensee shall evaluate and propose.

Concerning the alternate provision that the RCS be depressurized with an RCS vent of greater than or equal to 4.5 square inches:

We find that this should be confined only to MCDE 5. COLD SHUTDOWN,

. 0PS ARE NOT FILLED, and REFUELING OPERATIONS; MODE 6 HIGH WATER LEVEL and MODE 6 LOW WATER LEVEL. There are no safety analyses to succort

this type of operation in remaining MCDES 4 and 5. The proposed TS, j without this clarification, is nonconservative with respect to the I Licensing Basis. The licensee shall evaluate and propose.

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-1 We find no. safety evaluation in the Licensing Basis for the alternate use of an RCS vent of greater than or equal to 4.5 square inches.in the pec0csed T.S. The licensee shall evaluate and propose.

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T.S. SECTION 3/4.5 EMERGENCY CORE CCOLING SYSTEMS The operacility requirements from the McGuire Units 1 & 2 Licensing Basis FSAR are markedly different from those of the W Standarc Technical Specifications which have been adopted by the Licensee in his proposed T.S.

The Licensing Basis FSAR requirements are summarized under " General."

General FSAR Reference 8, page Q 212-47, Revision 25, item 212-75, describes the following Operator Instructions and Operator Actions During Shutdown.

"The sequences of events associated with shutdown will be described. The procedures associated with startup will be the same exca t they will be in reve.rse order. .The startup procedures are.not presented here to avoid .

unnecessary duplication. '

I Coerator Instructions Outing Shutdown A) At 1900 psig, the operator is instructed to manually block the

, automatic safety injection signal. This action disarms.the SI signals from the press,urizer pressure transmitters and from the steamline pressure transmitters. The SI signal on containment high "

pressure signal continues to be armed and will actuate safety injec-tion if the setpoint is exceeded. Manual safety injection actuation is also available. Also, at 1900.psig, the operator is instructed .

to close and gag UHI discharge valves. The UHI hydraulic pump and the gag motors for the UHI isolation valves are de-energized and tagged.

8) At 1000 psig, the operator closes the cold leg accumulator isolation valves. He then racks out, Iceks and tags the breakers for these valves. He also opens locks and tags the breakers for all safety injection pumos and all eut one charging pumo. At this time, one charging pumo and two resicual heat removal (RHR) pumps would ce j [gg, availaole for either automatic or manual SI actuation. _

C) At less than 400 psig and 350*F, the operator aligns the Resicual .-

Heat Removal System. The valves in the line from the RWST are closed. )

II Ooerator Actions During Shutdown A) Between 1900 psig and 1000 psig, the ECCS can either be actuated automatically by the high containment pressure signal or manually by the operator.

06/01/84 33A Revision A

B) Between 1000 psig and 400 psig, a portion of the ECCS can be actuated automatically (containment hign pressure signal) or manually by the coerator. The equipment that can be energi:ed are two RHR Oumos and one enarging pump. The operator would have to reinstitute power at the motor control centers or switchgear to the remaining safety injection pumps, enarging pump, and the accumulator isolation valves.

1. /L C) Selow 400 psig, the system is in the RHR cooling mode. The RHR system would have to be realigned as per plant startup procedure.

The operator would place all safeguards systems valves in the required positions for plant operation and place the safety injection, centrifugal charging, and residual heat removal pumps along with SI accumulator in ready and then manually actuate SI " l N-

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In resoonse to additional questions, the following information as proviced uncer FSAR reference 8, page Q 212-61, revision 28,. item 212.90(6.3); .

page Q 212-61a, revision 28, pages Q 212-61b, revision 29 and Q 212-61c, revision 29 "In spite of the low probability of occurrence and the fact that certain failure modes for pipe rupture do not exist during cooldown at an RCS pressure of 1000 psig, the following items have been incorporated into the station operating procedures:

1. At 100[0] psig, the operator will maintain pressure and proceeed to cool down the RCS to 425*F.

2. At 1000 psig and 425'F, the operator will close and lock out the accumulator isolation valves.

The above plant operating procedures will ensure that the accumulator isolation valves will not be locked out prior to about 2-1/2 hours after reactor shutdown for a cooldown rate of 50*F/hr.

A conservative analysis has defined that the peak clad temoerature resulting from a large break LOCA would be significantly less than tne 2200*F Acceptance Criteria limit using the ECCS equipment availacie l 2-1/2 hours after reactor shutdown.

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l The following assumptions were used in the analysis:

1. The RCS fluid is isothermal at a temperature of 425'F anc a pressure of 1000 psig.

2. The core and metai sensible heat acove 425*F has been removec.

3. The hot spot occurs at the core midplane.

4. The peak fuel heat generation during full pcwer operation of 12.38 kW/ft (l'02P. of 12.63 kW/ft) will be used to calculate adiabatic heatuo.

I 5. At 2-1/2 hours decay heat in conformance with Apoendix K of 10 CFR 50, the peak neat generation rate is 0.179 kW/ft.

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6. Two low head safety injection pumps and one high head charging pump are availacle from either manual Safety Injection actuation or automatic actuation by the containment Hi.-1 signal.

7. No liquid water is present in the reactor vessel at the end of blowcown.

8. A large cold leg break is considered.

For a postulated LOCA at the cooldown condition of 1000 psig, previous calculations show that the clad does not heat up above its initial temperature during blowdown. Proceeding from the end of blowdown and assuming adiabatic heatuo of the fuel and clad at the' hot spot, an increase of 446*F was calculated during the lower plenum refill transient of 89 seconds. During reflood, the core and downcomer water levels rise together until steam.generatic.n in the core becomes sufficient to innibit the reficoding rate. At that time, heat transfer from'the' clad at the .

hot spot to the steam boiloff and entrained water will commence. This neat removal process will continue as the water level in the core rises while the downcomer is being filled with safety injection water. The reflood transient was evaluated by considering two bounding cases:

1. Ocwncomer and core levels rise at the same rate. No cooling due to steam boiloff is. considered at the hot spot. Quenching of the hot spot occurs wnen the core water level reaches the core midplane.

2. Core reflooding is delayed until the S1 pumps have completely filled - _

the downcomer. No cooling due to steam boiloff is considered at the hot spot until the downcomer is filled. The full downcomer situation may then be compared with the results of the ECCS analysis in the SAR to obtain a bounding clad temoerature rise thereafter.

For Case 1 described above, the water level reached the core micolane A3.2 seconds after bottom of core recovery. The temperature rise curing reflood at the het scot from adiacatic heatup is 216*F, which results in a peak clad temperature of approximately 1086*F.

For Case 2, the delay due to downcomer filling is 54.4 sec. The corres-pending temperature rise at the hot spot form adiaoatic heatuo is 272*F, '"

which gives a not spot clad temperature of 1143*F.

The clad temperatures at the time when the downcomer has filled for the DECLG, CD = 0.6 submitted to satisfy 10 CFR 50.46 requirements are 1620*F and 1774*F at the 6.0 and 9.0 foot elevations, respectively.

Core flooding in the shutdown case under consideration will be more rapid from this point on due to less steam generation at the lower core power level in effect; decay heat input at any given elevation is less in the shutcown case. The comcination of more rapid reflooding anc lower power in :ne fuel insures that the clad temperature rise during reficod will be less for tne shutdown case than for the design basis case.

06/01/84 SS Revision A 1

Repeating the above calculation assuming the loss of a low head safety injection pump yields clad temperature of 1653*F and 1750*F for Cases 1 and 2, respectively. These results provide additional assurance that the peak clad tamoerature will not exceed 2200*F because, as statad above, in the shutdown case more raoid reflooding and lower power in the fuel insures that the clad tamparature rise during reflood wil be less than for the design basis case.

Based upon the analysis as presented above, it can be concluded that in the unlikely event of a LOCA at shutdown conditions, the peak clad temperature will be less limiting than that of the design base calculation.

N The response provided in Revision 28,[above] addressed the subject of operator actions and ECCS availability. Consistent with the information .

provided in Revision 28, a postulated LOCA in the RHR mode at 425 psig RCS pressure nas oeen assessed. The initial concitions would be reacned

',four hours after reactor shutdown. The integrity of the core after a postulated LC0A is assured if the top of the core remains covered by the resultant two-phase mixture. A conservative indication of time available for operator action is obtained by calculating the time required for the top of the core to just uncover. A calculation has been performed to confirm that margin for operator action does exist to prevent core uncovery.

This conclusion persists even under an assumption of ten minute delay for operator reaction time.

Assumptions: -

(a) The system pressure essentially reaches equilibrium with containment by the time the volume of water above the bottom of the hot legs is removed.

(b) Upper plenum fluid volume between the top of the core and bottom of hot legs is the only upper plenum fluic considered.

1 (c) Volume between tne core barrel and baffle is conservatively neglected.

(d) 120% of the ANS decay heat curve for four hours after shutdown is utilized.

f f Using the void fractions developed from the Yeh correlations and utili:ing a hydrostatic pressure balance, the height of the steam water mixture in the upper plenum was generated. Incorporating the plant geometry, tne

total liquid mass in the downcomer, core, and upper plenum was calculated, i.e., a mass-initial condition. Again by hydrostatic pressure balance, the height of liquid in the downcomer when the top of the core is just

' about to uncover was calculated. This information along with core volume is used .a develop a mass-final condition. That is, the mass is licufo l contained just before the core is uncovered. Utilizing the boil-off rate for the four hour time after shutdown, the time needed to evaporate a mass of mass-initial minus mass-final is calculated. This time was comcared to the ten minute assumption for operator reaction time.

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Utilizing the preceding approach, the time calculated to just initiate an uncovery of the core is 13 minutes. The conclusion is that even for the conservative method outlined above, there exists adequate margin to retain a safe core condition even in relation to a ten minute operator-response-tice assumption." j These operator requirements are verified, in general, by reference 12, SER Supplement 2 page 6.6-6.8 under " Emergency Core Cooling System - Performance Evaluation," and pages 7-1 and 7-2 under " Upper Head Injection Isolation Valves."

k i Additionally, the status of the ECCS systems from entry into the RHR MODE through cooldown, i.e., frem 425 psig/350*F through MODE 5 is clarified by the

'following extract from reference 11, Suppl. SER No 1, pages 5-1 and 5-2 wnich confirms continuance of the alignment at the end of MODE 3 425 psig/350*F /

througn both MODES 4 and 5. i 3

"5.2.2 Overoressure Protection In the Safety Evaluation Report we indicated a concern about the possibility of reactor vessel damage as a result of overpressurization when the reactor coolant system is water-solid during startup and shutdown. We have reviewed the applicant's system for overpressure protection when the reactor coolant system is water-solid. It consists of two separate trains each containing a '

power-coerated relief valve set to open when the system pressure reacnes 400 pounds per square inch gauge should an overpressure event occur. Each train contains an annunciator which sounds to alert the operator when plant conditions require enaoling of the water-solid overpressure protection system; enabling is performed manually, by turning key-1cck switch. The system is automatically disabled when plant conditions no longer require it; an annuciator sounds to indicate the system is no longer needed so that the operator may turn the key-lock to disable the system until needed. In addition, eacn train contains an'annuciator which sounds when the power-operated relief valve is coen, indicating an overpressure transient is in process.

Each power-operated relief valve is su plied with nitrogen from tne cold leg accumulators. No operator action is required in tne event of a transient.

The operator isolates the uoper head injection system, the cold leg ac:umulators, the safety injection pumps and one centrifugal charging pump before tne reactor

coolant system is cooled to 300 degrees Fahrenheit; only the remaining centrif-ugal charging pump could cause an overpressure transient as a result of inadver- i tent start with concomitant mass addition. The only other overpressure event would result from an inadvertent main coolant pumo start with the coolant in

'the secondary side of the steam generator hatter than that in the reactor coolant system. The applicant has shown that in neither case was 10 CFR Part 50, Apoendix G limit reached. For the latter case (that for main coolant pumo

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inacvertent start), the applicant assumed that the temperature of the fluid in the steam generator would exceed that in the reactor coolant system by no greater than 50 degrees Fahrenheit.

The staff requires that the technical specifications require that the reactor coolant system may not be cooled to temperatures lower than 300 degrees Fanren-hei.t without the ove pressure protection system enaDied, and unless totn 06/01/84 87 Revision A

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power-operated relief valve trains are operable, in order to assure suitable overpressure protection for the reactor coolant system when water-solid. In addition, the technical specifications will state that the temperature of the fliuid in the secondary side of the steam generator will not exceed the temp-erature of the fluid in the reactor coolant system by greater than 50 degrees i Fahrenheit wnen the reactor coolant system fluid temperature is less than 1 300 degrees Fahrenheit sinca the applicant's calculations did not assume differences greater than 50 degrees Fahrenheit.

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The applicant provided data to show that the power-operated relief valve opens within the time specified in the analyses.

The system meets the single failure criteria as only one of the two trains is required for overpressure mitigation. Means are provided to test and calibrate i

the system. It has been designed in accordance with the Institute of Electrical ard Electronics Engineers Standard 279-1971, ".Critecia for Protection Systems."

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This sy. stem meets the staff ret;uirements for an overpressure protection system with the reactor coolant system water-solid and is accootable. We consider this matter res 1ved.

The required status of the ECCS systems required by the existing Licensing Basis FSAR are briefly summarized: . .

Above 1900 psig (in MODES 1, 2, and 3): All ECCS systems are OPERABLE.

Between 1900 psig and 1000 psig/425*F; upper head injection isolation valves area closed and gagged, de-energized and tagged. Between 1000 psig/425* F and -

425 psig/350' F (in MODE 3): Upper head injection isolation valves remain closed and gagged and de-energized; cold leg accumulator isolation valves are closed and breakers racked out, 1 centrifugal and 1 reciprocating charging pump and 2 safety injection pumos are isolated, and rendered inoperable by i ocening and locking the related circuit breakers. Below 425 psig/350' (in

'dCDES 4 and 5) status of all ECCS systems remain unchanged, i.e., same (as for the preceding phase of 'dODE 3) with the exception that remaining ecuipment is re-aligned for RHR operation with the capability of re-alignment to ECCS.

[UHI, Cold Leg Accumulators, 1 cent. CP & 1 Recip. CP, and 2 SI pumos are i effectively electrically isolated.] RHR PORVs are rendered ocerable during water solid operation, belcw 3CO*F.

These requirements are substantially different from those of the W STS which -

the licensee has adopted for his facility contrary to his Licensing Sasis as disclosed in the FSAR and SER to the above references.

T.S. SECTICN 3/4 5.1 ACCUMULATORS / COLD LEG INJECTTON Item: APPLICABILITY MODE The Applicability Mode, given as MODES 1, 2 and 3* where 3" is 1000 psig, should be amenced to include 425'F; as 1000 psig/425'F. Reference the basis in the previous section entitled " General."

i l Since the proposed T.S. coes not contain this temperature constraint, it is i

non-conservative. A pressure of 1000 psig on the current Appencix G curve, l

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and T.S. temperature constraints, would permit an RCS temp of 557*F. The only =

available analysis in the Licensing Basis, see earlier under " General," shows I that cooling down to [1000 psig]/425*F is necassary to reduce the thermal burden on the ECCS so that the reduced ECCS capability can mitigate the consequences of a LOCA to 10 CFR 50.46 requirements; reference 8, pages Q 212-61, revision 28 and Q 212-61a, revision 23. The current T.S. is therefore non-conservative in this matter, and the licansee must evaluate and propose. Note; the " Footnote" Pressurizer Pressure above 1000 psig" also needs amendment. ,

7 l24,htem: 3.5.1.1.d.

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Nitrogen cover pressure is quoted at between 400 and 454 psig. The Licensing Basis FSAR, reference 4, page 1 of 5 revision 39 in Table 6.3.2-1 specifies a normal operating pressure of 427 psig. Making an allowance for channel error b and drift,should not this value be a higner sat point of accrox. 450 psig? The speci.fied set point values proposed in the T.S. of.400 to 454 psig.can.therefore ,

i give actual values wnich are lower than in the Licensing Sasis FSAR and ce non-conservative. The Licensee shall evaluate and propose.

O i Item 3.5.1.1.f Proposed The NRC proposes that an additional item limiting the range of actual water temoerature in the accumulator between 60-150*F in accordance with Licensing Basis FSAR referenca 29, Table 6.3.2-1 is necessary to confirm Safety Analysis l Limits for this accumulator. Its absence from the proposed T.S. rencers it potentially non-conservative. Further Item 4.5.1.1.1.a. concerning verifica-tion parameters should include Temperature of Accumulator Water. The licensee ,

shall evaluate and propose. .

I ACTION Items a and b require HOT SHUTDOWN generally, except for closed isolation valves. This may be too conservative - the licensee should review specific cases identified under 3.5.1.1.a-f and decide wnether HOT ShUTDCWN is nesassary instead of to 10C0 psig/425 F. Further, is there any conservative directicn of the error which may minimi:e his need to suspend operations at ;cwer, or allow him to operata at reduced levels. This licensee proposal may be unecassarily conservative. The licensee may evaluate and propose.

Item 4.5.1.1.c requires that once d

per 31 days wnen the RCS pressure is acove 2000 psig, it is verified that power to the isolation valve on tne Cola Leg ,.

Injection Accumulator is disconnected. What is the safety basis for this action, and where is it discussed in the Licensing Basis FSAR.

Item 4.5.1.1.1.d.1 requires that "At least onca per 18 months verify that each accumulator isolation valve opens automatically under each of the following corditions:

1) When an actual or a simulated RCS pressure signal exceeds the P-11 (Pressurf er Pressure 31ock of Safety Infection) Setpoint,"

We are not aware that this actually occurs; the licensee shall review and acvise of the related details within the FSAR on other licensing basis records.

This action is not described in FSAR reference 7, uncer Tacle 7.3.1-3 (1 of 2)

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and (2 of 2) revision 35, " Interlocks for ESFAS," nor in the related Logic Diagrams.

The LCOs of the Licensing Basis FSAR recuire that this Cold Leg Injection Accumulator be made operacle wnenever plant conditions exceed 1000 psig/425*F which is at a lower pressure than the current P-11 set point of 1955 psig; reference earlier T/S Section 3/4.5 under " General." This P-11 logic wnich would propose that this isolation valve is to be closed at RCS pressures between 1955 to 1000 psig is therefore non-conservative with respect to the Licensing Basis. The licensee shall evaluata and propose.-

12"L TMnsee shall verify that the set points for the relief valve on the Accumulators are included in the Inserv.fce Testing Program at the facility.

T.S Section 3/4.5.1.a (Pecoosed)

An additional T.S.' Section 15 proposed that provides.specifically for'the fact that " COLD LEG INJECTION ACC'JMULATOR ISOLATION VALVES" at " APPLICABLE CCNDI-TIONS" of MODE 3 (< 1000 psig/425'F), MODE 4 and M00E 5 would have a " LIMITING CONDITION OF CPERATICN" providing that "Each Cold Leg Injection Accumulator Isolation Valve is closed with circuit breakers opened, locked and tagged."

Accropriata Action Statements and Surveillance Procedures would be provided.

This is in accord with the LCOs of the Licensing Basis FSAR as described uncer earlier items T.S. 3/4.5, " General" and T.S. 3/4.5.1 of this review. Absence of this specific provision makes the proposed T.S. non-conservative. The licensee shall evaluate and propose.

T.S. Pace 3/4 5-3. UPPER HEAD INJECTION

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Item: APPLICABILITY MODE.

The Applicability Moce given as MODES 1, 2, anc 3" wnere

• signifies Pressuri:er Pressure acove.1900 psig, should be amended to include >425*F; as 1900 psig/>J25'F.

The FSAR does not incluce the temperature constraint explicitly at 1900 osig, thougn it is implicit in tnat the next lower boundary for change is 1000 psig/425*:

(Reference earlier Item: T.S. 3/4.5 under GENERAL]. Absent this concition, the relatec proposed T.S. is non-conservative. Appendix G curves (T.S.

Page 3/4 4-32) would allow RCS temperatures down to <300*F, and one of the reasons for isolating UHI below 1900 psig, incluces overpressure concerns at --

the reducing levels of temperature down to 425'F, reference 12, page 7-1. From his cetailed analysis, the licensee should evaluate and propose a lower limit to this temperature condition of >425'F.

Item 3.5.1.2.c Nitrogen cover pressure is specified as between 1206 and 1254 psig. The Licensing Basis FSAR, reference 29, page (1 of 5), revision 29 in Table 6.3.2-1 specifies a normal operating pressure of 1220-1290 psig with a minimum of 1220 psig. Making an allowance for channel error and drift, should not T.S. setpoints be higher (at say 1240-1300 psig]. The specified minimum set point values in the proposea T.S. of 1206 would therefore require lower pressure in the RCS before actuation and is therefore non-conservative. The licensee shail evaluate and propose.

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1 3 ' Item 3.5.1.2.d: Proposed.

It is proposed that an additional item limiting the range of actual water temperatures in the accumulator to between 70 and 100*F in accordance with reference 29, Page (1 of 5), revisien 39, in Table 6.3.2.1 is necessary to confirm the Safety Analysis Limits for the UHI Accumulator. It is also pro-posed that it be added as an additional surveillance element to item 4.5.1.2.a. '

Its absence from the proposed T.S. renders it potentially non-conservative with '

( respect to the Licensing Basis. The licensee shall evaluate and propose.

Action Items a & b require HOT STANOBY, generally, except for closed isolation valves, followed by HOT SHUTDCWN. This may be too conservative - the licensee should review specifically each of the Operacility items b, e and proposed d, and decide wnether HOT STANOBY leading ultimately to HOT SHUTCOWN is necessary.

Further, he should assess if either boundary value, upper or lower, can be conservative, anc oy how much, and evaluate whether ne shoulc taxe,an ACT!CN STATEMENT under " conservative" conditions. The licensee may eva10 ate and propose. ,

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The licensee shall verify that the relief valve set point on the Accumulator is included in the In Service Testing Program at the facility. j T.S. Section 3/4.5.1.b (Proposed)

An additional T.S. item is proposed tilat provides specifically for the fact that " UPPER HEAD INJECTION SYSTEM ISOLATION VALVES" at APPLICABLE CONDITIONS of MCDE 3 (< 1900 psig and > 425'F), MODE 4 and MODE 5, wouTd have a " LIMITING CCNDITICN OF OPERATION" providing that "Each upper head injection system isola-tion valve" is closed and gagged. The UHI hydraulic pump and the gag motors for the UHI isolation values are de-energized and tagged. Appropriate Action Statements and Surveillance Procedures would be provided. This in accorcance with the LCOs of the Licensing Basis FSAR as described in earlier items T.S. 3/4.5, " GENERAL" and T.S. 3/4.5.1 of this review.

Absence of this soecific crovision makes the current T.S. non-conservative ita respect to the Licensing Basis. The licensee shall evaluata anc proocse.

T.S. Section 3/4.5.2 ECC SUBSYSTEMS -Tave 2 350'F The title snould be amended to read as: -

ECCS SUBSYSTEMS - PRESSURIZER PRESSURE > 1000 psig/RCS Tavg3425'F The Operacility requirements of 2 full trains of ECCS equipment remains unenangec.

Absence of the pressure /temoerature condition in the proposed T.S. is not in accordance with Safety Analysis Limits. Its acsence permits high pressure pumc oceration at lower pressures and temperatures with potential infringement of related safety criteria. Related safety criteria have not been well definec, or docxetec, but are apoarently considerations of Low Temperature Overpressure Drotection of tne RCS under these and related Accident circumstances inclucing inadvertent operation of ECCS pumos. This diversion from the Safety Analysis ,

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I Limits of the Licensing Basis FSAR must, therefore be considered non-conservative If'( and the licenseee shall evaluate and propose.

Item 4.5.2.h.: concerning flow balance tests in the ECCS system. The licensee snall provide the bases for the flow distributions specified and furtner advise how they might meet minimum flow conditions to intact loops creeng Accident f-Occurrences.

T.S. Section 3/4.5.2.A Procosed A proposed new Section which would be titled: ECCS Subsystem - Applicability between 1000 psig/425'F and 425 psig/350*F.

This would provide for: One ECCS subsystem comprising the following shall be OPERABLE:

a. One OPERIBLE centrifugal charging pumo,# , .

b. One OPERABLE RHR heat exchanger,

c. One OPERABLE RHR pump, and

d. An OPERABLE flow path.

Also,' one ECCS subsystem comprising the following shall also be OPERABLE

b. One OPERABLE RHR heat exchanger,

c. One OPERABLE RHR pump, and ,

d. An OPERABLE flow path All creakers for all safety injection pumps and all but the one operable centrifugal charging pumo are opened, ' locked and tagged (reference earlier information).

l As explained in the previous section, limited operation of the higner pressure l pumps between 1000 psig/425'F and 425 psig/350*F acparently provides Low

, Temperature Overpressure Protection (LTOP). The proposed T.S. requires all j CI and SI pumos to be available during these conditons and is therefore ~,

i non-conservative with respect to the Licensing Basis and particularly in rescect of Overpressure Protection. The licensee shall evaluate and propose, and in so doing provide the analyses and evaluation which required constrained operacility

, of the higher pressure pumps in this operating phase, in his Licensing Basis j FSAR.

T.S. Sectior 3/4.5.3 ECOS Subsvstem - Tava 1 350*F

.This title should be amended to read ECCS Subsystems - 425 psig/350*F to COLD j SHUTDOWN t

l The current T.S. provices no pressure condition on the temperature of 350*F, and Appendix G Limit curves of proposed T.S. Page 3/4 4-32 would permit " maximum 06/01/B4 92 Revision A l

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RCS pressures" of 2485 psig under these circumstances. Also the procosed T.S.

alignment eliminates safety injection and charging pump capacity. There is no available evaluation of the capability of the reducac ECCS system to satisfac-torily mitigate the consequencas of a Small Break or Large Break LOCA from 2485 psig/350*F as is provided for the values of 425 psig/350*F within the Licensing Sasis as described earlier under T.S. 3/4.5 Item: GENERAL. Our evaluation is that the absence of this pressure condition is non-conservative, and especially with respect to the Safety Analysis Limits of the Licensing Basis. The Licensee shall evaluate and propose.

The proposed limit at COLD SHUTD0' sin MODE 5 is conditioned by the fact that Refueling is a condition of a vented vessel with Reactor Vessel Bolts unten-sioned, and non-ECCS alignments are proposed to deal with related events.

R,eference 8 pages Q212-56 revision 25 under the Titles of Case 1 and Case 2 and page Q 212-57, revision 25, uncer the Title of Case 3. Overpressure Protection

. also, wnich'is a principal determinant of alignment, also ceases with unten-sioning the Reactor Vessel bolts for refueling.

The proposed T.S. under this Section requires a minimum of one only ECCS subsystem comprising

a. One Operable Centrifugal Charging Pump (CCP)'

b. One Operable RHR Heat Exchanger

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c. One Operable RHR Pump

d. An Operable Flow Patn There are no Safety Analyses or Evaluations of one only ECCS suosystem allowing for a single active failure in one cnly train. This proposition is therefore non-conservative with respect to the Licensing Basis FSAR. The Licensee sna11 evaluate and prooose.

This T.S. coes not disallow the accitional CCP and 2 Safety Injection Pumos (SIPS) from 350*F down to 200'. This again is non-conservative witn resoect '

to the LCOs of the Licensing Basis FSAR wnich allows only one (1) CCP, anc the I remainder f.e. , one (1) CCP and any otner reciprocating enarging ::umo anc 2 SIPS ,

are to be electrically isolated against inadvertent operation. This procosed T.S. is again non-conservative in respect of overpressure crotection wnen ecm- j pared with the current Licensing Sasis. The licensee shall evaluate and J propose.

The proposed T.S. allows one (1) CCP and one (1) SIP whenever the RCS temp is less than 300*F. The LCO of the Licensing Basis FSAR allows only one (1) CCP because of OVEMESSURE PROTECTION; reference earlier information under earlier T.S. Section 3/4.5. Item: " General". The proposed T.S. is therefore non-conservative with respect to the Licensing Basis. The Itcensee shall evaluate ano propose. j The LCOs of the Licensing Basis FSAR recuire the same coeracility of ECCS ecuipment as is required for TS 3/4 5.2A Proposed. So that in addition to:

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One ECCS subsystem comprising the following shall be OPERABLE:

a. One OPERABLE centrifugal charging pump, L

b. One CPERA3LE RHR heat exchanger,

c. One OPERASLE RHR pump, and

d. An OPERABLE flow path H

which is the same as for. the proposed T.S. , it is also required that:

One ECCS subsystem comorising the following shall also be OPERABLE:

b.. One OPERABLE RHR heat exchanger,

c. One OPERABLE RHR pump, ana

, d. An OPERABLE flow path.

Additionally, that all breakers for all safety injection pumos and all but the one operable centrifugal charging pump are opened, locked and tagged.

(reference earlier information) The proposed T S. is therefore less conserva-ttve than the Licensing Basis FSAR by being deficient in ECCS total pumoing-capacity, and excessive in available high pressure pumping capacity so "-

infringing LTOP. The licensee shall evaluate and propose. .

Additionally the Licensing Basis requires that och of these subsystems be independent and receive power from two (2) redundant Emergency Buses and Power Sources.- The absence of any such provision in the proposed T.S. makes it non-conservative with respect to the Licensing Basis. The Licensee shall evaluate and propose.

T/S Section 3/A.5.4 BORON INJECTION SYSTEM / BORON INJECTICN TANK.

Item: APPLICABILTY MCCES 1, 2, and 3 with the current crecosed T.S. snoulc ce enanged to include MCOE 4 in accorcance with the Licensing Basis FSAR .ni:n evaluates MSLS and LCCA events down to and including this MCCE. Acoption of the Licensing Basis FSAR mode of boration control may eliminate this need.

. With proposed T.S., however, the absence of the BIT tank in Mode 4 must'oe -

considered non-conservative. The licensee should evaluata and propose.

! Item: The ACTION Statement should be clarified to include [ ] that in One

! event of inoperablity of the BIT tank, the RCS be borated to [a baron concentra-I tion which will give] a SHUTDOWN margin of 1% delta k/k at 200*F.

The li..nsee shall clearly indicate, that this item is not applicaele to Unit 2 by reason of a recent SER from NRC.

Comment: Since BIT concentrations of only 2000 ppm, only are now requitec, and only 900 gallons are involved compared with 372,100 gallons in the R.W.S.T. is not the proposed ACTION statement to ultimately place the plant in HOT SHUTDC'aN overly conservative; if minimum volumetric requirements are necessary, can 06/01/84 94 Revision A T-'-"- -

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additional provision be made in the RWST. The licensee may evaluate and propose.

-T.S. Sectien 3/4.5.5 REFUELING WATER STORAGE TANK Item: APPLICABLITY MCOES 1, 2, 3, 4.

The current MODES 1, 2, 3 and 4 which includes an LCO for 372,100 gallons must be extended to MODE 5 and MODE 6 (limited) to meet the FSAR requirements in reference C, pages Q 212-57 and 58, revision 25, item: Case 3: [when] The RCS is depressurized and vented with the air in the steam' generator tubes, with the reactor vessel head on, and tensioned - and later with open relief paths between the head and the reactor vessel cavity and refueling canal. The single failure of an RHR/RCS Isolation valve is resolved by the expected Operability of the RWST providing 5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> of injection flow. The recovery description also

• means that tne RWST must be availaole in MCOE 6 until the vessel heaa is removed a'nd the refueling canal is filled to its spec 1'fied level. It must also be .

available at termination of core alterations - in Mode 6, when drainage of the refueling canal commences until the Reactor Vessel Head is tensioned, when the RCS then moves into MCDE 5. The proposed T.S. is non-conservative with respect to the Licensing Basis. The licensee shall evaluate and propose.

Action Statement: The proposed ACTION should be modified [ ] as follows:

Witn the RWST Inoperabl'e, restore the tank to O'PERABLE status within l' hour, or be in at least HOT STANOBY [and borated to a boron concentration which will give a shut down margin of 1% delta k/k at 200*F and a minimum of 2000 ppm]

within [the next] 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

The Licensing Basis FSAR requires Safety Injection of 2000 ppm Boron to mitigate the nuclear power consequences of any accidents which may initiate during tnis period; if the RWST is not available, then Boron Concentration in the RCS snould be increased to the level required to mitigate any potential return of nuclear power. The croposed T.S. accears nonconservative.

The licensee shall evaluate and precose and in so coing he should evaluate each of the Operability recuirements separately to determine if COLD SHUTCC'aN is requirec. for each INOPERABILITY REQUIREMENT, or whether alternate mitigating Actions are possible.

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T.S. Section 3/4.7 PLANT SYSTEMS T.S. Pace 3/4 7-1: SAFETY VALVES The proposed T.S. requires that:

3.7.1.1 All main steam line Code safety valves associated with each steam generatorshallbeOPERABLEwithliftsettingsasspe(fiedinTable3.7-3.

APPLICABILITY: MODES 1, 2, and 3.

ACTION: ,

s. With four reactor coolant loops and associated steam generators in operation and with one or more main steam line code safety valves .

inoperaole, operation in MODES 1, 2, and 3 may proceed provided, that within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />, either the inoperaole valve is restored to OPERABLE status or the Power Range Neutron Flux High Trip Setooint is reduced per Taole 3.7-1; otherwise, be in at least HOT STANCBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />,

b. With three reactor coolant loops and associated steam generators in operation and with one or more main steam line code safety valves associated with an operating loop inoperable, operation in MODES 1, 2, and 3 may proceed provided, that within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />, either the inocerable valve is restored to OPERABLE status or the Power Range Neutron Flux High Trip Setpoint is reduced per Table 3.7-2; otherwise, be in at least HOT STANOBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTOOWN with*the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

Our concerns in this section are parallel to those in our review under T.S.

Section 3/4.4.2 SAFETY VALVES.

Failure of Steam Generator Code Safety Valves infringe basic safety criteria l for Reactor Protection through its impact on SG/RCS system resconse under Condition II, III, and IV occurrences. It also affects the integrity of the Primary Containment Boundary.

We do not find an adecuate consideration of the alternate type of Safety Valve Failure that can occur, and their related significance, ucon the action state-ments proposed.

How sure is the Licensee that inadequacy to meet the very limited single operacility requirement of the T.S. does not represent an intermittent problem leading to early opening of valves, failure to close, or failure to open under tt severe concitions of Transient and Accident Events.

We find the proposed T.S. inadequate in its representation of operability, or lack there of, for these Safe'ty Valves. Consequently, without a requirement that they all be oceraole in MODES 1, 2, 3, and 4, with a further requirement 06/01/84 96 Revision A

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t to be in cold shutdown in the event of failure, there of, we must consider the proposed T.S. non-conservative. The Licensee shall evaluate and propose.

T.S. Pace 3/4 7-4: AUXILIARY FEECWATER SYSTEMS Itam: APPLICA8ILITY MODES 1, 2 and 3 in the proposed T.S. should be expanded to MODES 4 and 5 in accordance with our review under Table 3.3-3 ESFAS INSTRUMEN-TATION, Items 7 a, b, c, d, e, and f. The conclusions from that review are:

The proposed T.S. items are generally non-conservative with respect to the Licensing Basis. The licensee shall evaluate and propose. _

!8eg Item 3.7.1.2.b. The licensee has deleted OPERABILITY requirements for the g Steam-Turoine driven auxiliary feecwater pump at steam pressures of less than i 900 psig. This is not in accord with current Accident Analyses and no justifi-cation has been provided: Reference 15, Recommendation GL-3, recuires the Steam-Tureine AFW pump in the event of complete loss of AC pcwer for a peried of 2 hrs and beyond. This will require cperability down to the lowest pres-sures for wnich the Turcine is provided as described in reference 22, Table 10.4.7-6 wnere the range of operating pressures provided for is frem 110 psig to 1205 psig. This will also provide for operabilty down to and f including MODES 4 (and availabiilty from MODE 5) to cover licensing require-ments discussed elsewhere under Taole 3.3-3, ESFAS INSTRUMENTATION, Items 7a through f. .

We note two principal features relating to the service conditions of the Turcine Driven Feedwater Pumps: -

a. They are supplied with steam from two steam generators from main steam lines after the flow restriction orifices at outlets from the Steam Generato.rs.

b. They would normally be expected to perform early in the transient and continue to function to design ficw requirements througncut the Occurrence.

t The licensee should explain now tne crocosed TS ensures that the Turcine Criven  ;

pumo maintains its ficw performance required by Acciden. Analysis *nen steam line pressures could drop sucstantially below the Steam Generator Pressures due to presence of the SG flow restrictions and until main steam isolation valves "

are isolated on steam line pressure of less than 565 psig (< provides for channel crift and errors).

The licensee shall evaluate the above comments and propose tecnnical specifi-cations which will ensure operability of the Turoine-C .sen AFW Pu=o over tne range of conditions expected from Design Basis Accident Analysis, and other less bounding events, down to and including MODE 4 as discussec in the Licencing Basis.

In his evaluation, the ifcensee should advise if Item le of Table 3.3-5 ESFAS

)

INSTRUMENTAT!ON, Steam Line-Pressure Low is derived from steam line sensces and after tne SG orifices, or if it is taken from pressure, sensers on the Steam Generator. The licensee should tnen aavise what has been used in assessing Steam Generator Pressure Response and Turbine Driven AFW pumo resconse in the 06/01/S4 37 Revisica A

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Condition III and especially Condition IV Occurrences of the Licensing Basis, and if the existing Accident Analyses remain valid.

Item 4.7.1.2: SURVEILLANCE REQUIREMENTS The Technical Specifications, page T.S. 3/4 7-4 requires each motor driven (MD)

AFW pumo to supply 450 pgm at greater than or equal to 1210 psig. This is at entrance to the Steam Generators according to the T.S. Basis on T.S.

page B 3/4 7-2.

However, we note that the FSAR Accident Evaluation; reference 7, section 15.4.2.2.2, and the description of the AFW system in reference 5, refer to a total supply of 450 gpm from MDAFW pumps to three intact steam generators.

Further, this is parallel with a description in the Accioent Analysis on page 15.4 - 13.a (Revision 38) in which the MDAFW pumo headered to two intact

' steam gene,rators supplies 170 gpm each whilst the one headered to the faulted Steam Generator suppies 110 gpm to tne intact steam generator.

The SER supplement, reference 14, page 10-2 requires that the licensee confirm the capability of each of the Motor Driven and Turbine Driven AFW Pump systems to meet the flow distribution requirements of that particular Safety Evaluation Report, with a faulted steam generator associated with the ruptured main feedline and a second steam generator (SG) faulted with a failed open code Safety Valve or SG PORV, and botn these SGs supply the Turbine Driven AFW pumo. The Licensee committed to establish and verify by test, the valve throttle positions neces-sary to acnieve this, during the initial startup test programs. .-

In addition, under SER supplement, reference 15, page 22-15, under the title of Recommendation GS-6 the licensee agreed to propose Technical Specifications to assure tnat prior to plant startuo folicwing an extended shudown, a flow l test would be performed to verify the normal flowpath from the primary AFW system to the steam generator. The flow test should be concucted with AFW system valves in their normal alignment.

At this time, ne do not see a procesed T.S. wnich ensures that the recuirec succivision of flow between 3 intact and 1 faulted steam generator, anc 2 Intact and 2 "Faultec" Steam Generators associatec with the Turcine-Oriven AFW Pump, recuired by the Licensing 3 asis is acnieved, and we do not see any

=^

test period recommended such as following an extended cold snutdown to ensure that the recuired flow division is maintained in an acceptable manner. At tnis l time we must concluce that the current T.S. is nonconservative in respect to the '

Licensing Basis. The licensee shall evaluate and propose.

T.S. Page 3/4 7-Ec Procesed: CONDENSATE STORAGE TANK SYSTEMS l

It is proposed that a new item be added to the Technical Specifications to the above title and to include an LCO providing "The Condensate Storage Tank System (CTS) comprising available usable storage from the upper surge tank, auxiliary feedwater condensate storage tank and condenser hot well snall be operacle ith a contained water volume of at least 175,000 gallons of water.

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APPLICABILITY MODES proposed are 1, 2 and 3, with lesser volumes required in MOCES 4 and 5.

ACTICN STATEMENT should include a provision that, with the condensate storage tank inoperacle, within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> either

a. Restore the CST to OPERABLE status or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in HOT SHUTDOWN within the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />, or Demonstrate the OPERABILITY of the Nuclear Service Water System and Standby Nuclear Source Water Pond (alternate water source) as a backup supply, and align to the auxiliary feedwater pumos, and restore the condensate storage tank to OPERABLE status within 7 days, or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in HOT SHUTDOWN within th,e following 6 nours.

SURVEILLANCE REQUIREMENTS should include

a. The condensate storage tank system shall be demonstrated OPERABLE at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> by appropriate measures when the tank is the supply source for the auxiliary feedwater pumps,

b. The Nuclear Service Water System and Standby Nuclear Source Water Pond shall be demonstrated OPERA 8LE at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> by appropriate measures. ,

Additionally, an evaluation of and provision will need to be made concerning

- potential loss of AFW supplies during loss of suction and change-over to alternate AFW sources.

f The safety basis for these requirements are

a. Our earlier review uncer TS. Table 3.3-5 Items 7a anc 7b show that wnereas all safety evaluations involving AFW scoply have assumed a Safety Analysis Limit of 61 sec. response time, this is only availacie from nonsafety relatea sater sources. 'Further, that the safety related supD1y from the Nuclear Service Water Pond may take an extra 15 secs which is substantially non-conservative in resoect of the '

related safety analysis.

Therefore, at this time, until the licensee has evaluated our concerns anc mace ac:eptable proposals, the NRC will require technical specifications on this nen safety-related water storage of the acove nature. The proposed T.S. are nonconservative with respect to Regulatory Requirements. The licensee shall evaluate and propose.

T.S. Pace 3/4 7-8: MAIN STEAM ISOLATION VALVES Item 3.7.1.4 The procosed T.S. provides that: "each main steam line isolation valve (MSLIV) shall be OPERABLE with APPLICABILITY MCDES 1, 2, and 3.

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The requirements within the Licensing Basis for Main Steam Line Isolation are discussed in this review under Table 3.3-4, Item 4 The Licesing Basis does require operability .m MODE 4, in addition to MODES 1, 2, and 3 already provided.

We also note that the Main Steam Isolation Valves are Containment Isolation Valves as defined by 10 CFR 50 App. A Criterion 57 " Closed System Isolation" and the Licensing Sasis FSAR under reference 4 Table 6.2.4-1 (sheet 7 of 11)

Revision 4 and that Primary Containment Integrity is required in MODES 1, 2, 3, and 4 according to proposed T.S. Section 3/4.6.1, T.S. Page 3/4 6-1.

The proposed T.S. is non-conservative with respect to the Licensing Basis; the Licensee shall evaluate and propose.

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' T.S. Pace 3/4 7-Sa Proeosed: STEAM GENERATOR PCWER OPERATED RELIEF VALVES (SG PORVs) . .

The proposed T.S. does 'not ine'1ude these valves whic'h are required to enable the plant to be cooled down under natural circulation conditions (under Loss of Offiste Power]. The Licensing Basis requirement for this is described in SER Supp No. 4 reference 14 page 5-7.

The minimum number of valves required for natural circulation has not been established in the Licensing Basis. Reference 15, page 15.2-28, revision 15, .

under section 15.2.9.2 discusses natural circulation as verified by Table 15.2.9-1 which is at a maximum of. 4%. This review; under earlier Table 2.2-1 Item 18b, shows how the existing Control Logic can place this plant into a natural circulation Occurrence, without reactor trip at a nominal power level i

of 12E Rated, and the review under Table 3.3-1 under Item: .Concerning Prescribed Values for % Rated Thermal Power DURING START UP (MODE 1) AND POWER OPERATION 4

(MODE 2) shows how the resulting residual nuclear power levels could actually be the order of 20%. Therefore, in addition to the evaluation required of the Licensee to meet those circumstances as described therein, he shall consider

, the consequences of the very limited SG PORVs capacity currently availacle to

[ meet this situatien. The Licensing Basis FSAR. reference 9, page 10.1-2, revision 8, para 3 shows a capacity of only 10% [without single failure).

This means that in addition to the potential inability of the RCS to provide the requisite cooling capacity under natural circulation for a nominal 10%,

and potential 20%, power level, the SG PORV capacity is insufficient in the event of a single failure (of 4 available) for nominal conditions, and severely ,.

under capacity for a possible 20% power level. At this time, until further evaluation has been completed, the Licensee should ensure, within the T.S., a potential atmospheric relieving capacity of 20%, allowing for a single failure.

This should include all his SG PORVs, plus elements of the additionally availacle 45% (of full load main steam flow to atmosphere) described under reference 22, page 10.1-2, revision 8, para 3, if they can be available under Loss of Offsite Power. An approcriate Action Statement should be provided. If the additional l ' atmospheric relief is not available on LOOP, the Licensee must further evaluate and propose necessary corrective actions.

The current omission of SG PORVs from the T.S. is non-conservative with respect to the Licensing Basis. The current omission of relieving capacity additional 06/01/84 100 Revision A

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to the SG PORVs is contrary to Regulatory Requirements which have been excluded from ne Licensing Basis. Tne Licensee shall evaluate and propose.

T.S. Section 3/4.7.3: CCMPONENT COOLING WATER SYSTEM The proposed T.S. requires that:

3.7.3 At least two independent component cooling water loops shall be OPERABLE.

APPLICABILITY: MODES 1, 2, 3, 4 ACTION:

With only one component cooling water loop OPERABLE, restore at least two loops to OPERABLE status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or be in at least HOT STANC3Y within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.,

The SER for the plant under reference 10, summarizes the following Licensing Basis for the Component Cooling System: ,

9.2.4 Comconent Coolino System The component cooling system provides cooling water to selected nuclear auxiliary components during normal plant operation and cooling water to safety-related systems during postulated accidents.

The component cooling system is designed to: (1) remove residual and .

sensible heat from the reactor coolant system via the residual heat ~

removal system during shutdown; (2) cool the letdown flow to the chemical.

and volume control system during power operation.; (3) cool the spent fuel pool water; and (4) provide cooling to dissipate waste heat from various primary station components during normal operation and postulated accicent conditions. Active system comoonents necessary for safe plant snutcown are designed to incluce at least 100 percent redundancy. The cocoonent cooling water for eacn unit includes two comoonent cooling heat excnangers, four component cooling pumps and a split-volume comoonent cooling surge tank. Two pumos anc one heat exchanger per unit provide the necessary cooling water for normal operation, cooldown, refueling, anc postulatec accidents. The remaining pumps and heat exchangers serve as stancby. An -

assured supply of makeup is provided from the nuclear service water system to each reduncant loop.

The component cooling water system is designed to seismic Category I requirements, except for certain branches to non-essential equipment.

The component cooling water pumps are powerec by redundant emergency buses. The portion of the component cooling water system serving the residual heat removal system meets the single failure crite 'on for active components. ,

I Based on our review, we conclude tnat the component cooling system cesign is in conformance with the requirements of General Design Criterion la 06/01/8a 101 Revision A

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of Appendix A to 10 CFR Part 50 regarding the capability of the system to transfer heat from systems and comoonents important to safety to an "

ultimate heat sink and provisicns of suitable redundancy for safe cool-down. We further conclude that the system design meets the requirements l of General Design Criteria 45 and 46 of Apoenaix A to 10 CFR Part 50  ;

regarding system design that allows performance of periodic inspections g6 and testing. We cencluce that the component cooling water system is ,

acceptable. [

I Detailed reference to Operability and Operating requirements in the Licensing Basis in MODES S and 6 can be found in reference 22, page 92-17 and Component Cooling System.

The proposed T.S. completely ignores, without'any evaluation, the Licensing Basis requirement for this system in MODES 5 & 6. The current T.S. are non- ,

conservative with respect to the. Licensing Sasis. The Licensee shall evaluate

• and propose. g This T.S. is a prime example of a Standard Technical Specification wnich completely ignores the Licensing Basis for all Nuclear Power Plants. This i reflects a very serious Safety Issue for all standard T.S. and wnich cannot I await an extended " Generic" Resolution.

I T.S. Section 3/4.7.4 NUCLEAR SERVICE WATER SYSTEM l APPLICABILITY MODES proposed are 1, 2, 3, 4. These should be extended to MODES 5 and 6.

Within the Licensing Basis FSAR, reference 6, [vol 8] page 9.2-5, "The Nuclear Service Waste System (NSWS) is designed to meet single failure criteria with two redundant channels [per unit] to serve components essential for safe station shutcown." The equipment requiring NSWS also includes all RPS and ESFS systems, many of wnien are necessary in MCDES 5 and 6 to the above redun-j dancy and single f ailure critaria.

' Etamoles include: MODE 5 is required to service AFW alternate cooling recuire-ments in event of a fail-closed RHR/RCS isolation valve in the RHR line, and in MCDES 5 and 6 it is neeced to service necessary redundant RHR Trains.

Reference our related evaluations in this review concerning RHR coerability requirements in MCOES 5 and 6. ..

The procosed T.S. is nonconservative with respect to the Licensing Basis. The licensee shall evaluate and propose.

T.S. Section 3/4.7.5 STAN08Y NUCLEAR SERVICE WATER PCND (SNSWP) l

! Item 3.7.5.b, an LCO, should be amended to read that the nuclear service water

( pond shall be operable with "an average water temperature of not less than 70*F or greater tnan 94*F l ....in the intake structure" l 102 Revision A 05/01/S4 l

The Licensing Basis FSAR, reference 6, page 9.2 - 12(a), revision 39, item 39, provides for an allowable maximum of 94* which meets both maximum allowable temperatures for all Safety Related Ccmponents including NPSH requirements (reference 6, page 9.2-13, last para).

An average water temocrature of 70'F has been selected by RSB as a potential design basis for Condition II, III and IV occurrences. The licensee has pro-vided little information on the range of AFW temperatures used in his analyses and the related sensitivity of results to AFV temperature variations. In the Major Rupture of A Main Feedline, reference 7, page 15.4 - 13, it is stated that a "relatively cold (120*F) AFW temperature was used (after purging the feedwater lines)." " Excessive Heat Removal" analyses in reference 7, page 15.2 - 29, uses a " conservatively low feedwater temoerature of 70*F."

We note that reference 6, page 9.2-13, revision 39, item 8 discusses ice formation on the surface of the pond which would imply near freezing temcer-atures for water supp'ly. At this time, we have no . record of any Safety -

Analysis being undertaken at such low inlet temperatures and on this basis we must consider any such low value as non-conservative.

The licensee will advise the range of AFW temoeratures used in Condition II, III and IV events, their sensitivity to AFW temperature values, and from this his bases for setting any alternate values proposed to the water temoeratures in the stancby nuclear service water pond. The proposed TS maximum value of 78'F is conservative with respective to certain Accident Analyses; the lack of a minimum temperature of 70*F including possible near-freezing temperatures must be considered as nonconservative in respect of certain events. The Licensee shall evaluate and propose. ,,

APPLICABLE MODES: The system is required in all MCDES 1, 2, 3, 4, 5, & 5 to handle heat rejection requirements as the ultimate heat sink. The licensees proposal to limit this to MODES 1, 2, 3 and 4, is nonconservative witn rescect to the Licensing Basis. The licensee shall evaluate and propose.

Ilr Reference 6, page 9.2-13, revision 39, states that "In the event of solic layer of ice" forms on the SNSWP, the ocerating train (of the Nuclear Servica Water [NSW) systenQ is manually aligned to the SNSWP. The Licensee snall provice the Safety Related reason for this action and advise if this coerster action conflicts with the Response Times proposed under Table 3.3-5. Given a ,,

Safety Related reason, surveillance requirements ensuring thi,s action snoulc be included under either T.S. Section 3/4.7.5 NSWS or this particular T.S.

Section 3/a.7.5 STANDBY NSWP. Absent this surveillance recuirement on a Safety Related Issue, the proposed T.S. would be non-conservative. The Licensee shall evaluate and propose.

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T.S. Section 3/4.9 REFUELING OPERATICNS e T. S. Item 3/4 9.1 BORCN CONCENTRATION Additional LCOs are necessary to meet the requirements of reference 8, page 15.2 - 14, revision 10 concerning Accident Evaluation for Section 15.2.4, Uncontrolled Boron Dilution. The baron dilution analyses of this reference 7, provides that, during refueling: I

a. " Amin [mumwatervolumeintheReactorCoolantSystemisconsidered. -

This corresponds to the volume necessary to fill the reactor vessel above the no::les to ensure mixing via the residual heat removal loop." -

L

~

, 'D. Neutron sources are installed in the core and the sourca range- ,

. ~ detectors 6utside the reactor vessel are active and provide an- l audible count rate. ,

c. A high flow alarm at the discharge of the CVCS (from flow element INVFE 5630) is active providing an alarm to the operator wnen the .

flow rate from the charging pumps exceeds 175 gpm. ,

I

d. The charging pumps are inoperative, j Aoditionally, an appropriate condition which must be attached to'a) above is ~

that any such minimum volume should be such that the level of water in o* above the loop provide acceptable flow, including NPSM conditions, at inlet to the -

RHR pumps.

These conditions are appropriate LCO's to 10 CFR 50.36; their current absence from the T.S. for this MODE is a non-conservative situation in respect of the Licensing Basis, and the Licensee shall evaluate and propose.

1 The current SER, Supolement No.1, reference 11,15-1, provides that: s I

"During refueling the acclicant has committed to isolate all sources of unoorated water connected to the primary system refueling / canal / spent f fuel. I l ,e

'ae do note that Surveillance Requirement T.S. 4.9.1.3 does provide for verifying tnat valve No. INV-250 is closed, under administrative control in succort of this. However we do note that according to reference 7, page 15.2-15, item Q 212-58, this valve INV-250 is to be locked closed during refueling. The current position could be non-conservative if the valve is not specifically locked under the proposed administrative control. Also notice, that reference 7, page 15.2 - 14, revision 10 states that:

"The other two paths are through 2 inch lines, one of which leads to the volume control tank with the other bypassing this tank. These lines contain flow control valves INV171A and INV175A respectively."

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Why are T.S.s not aoplied to the closure of these valves also? The preposed  ;

T.S. may be nonconservative with respect to the Licensing Basis. The licensee i shall evaluate and propose,

_\

We also note an apparent non-conservative ciscrepancy between the basis for the specified reactivity condition of "a k, , of 0.95 or less" with ut any specification of the position of movable ccntrol assemblies. We also note the need to add, according to reference 7, page 15.2-14, revision 10, that the baron concentration is to give a shutdown margin of at least 5 per cent delta k with all the rod cluster control assemblies out. The additional requirement uncerlinea should be a part of the LCO for this T.S. f ten. Without this pro-vision in the proposed T.5, it could be interpretad as non-conservative in respect of the Safety Analysis Limits for the plant. The licensee shall evaluate and propose.

In the L'icensing 3 asis FSAR. reference-8, page Q 212-24, item 212.57, it is -

required that the reac*cr makeup water pumps shall be removed from the loacs supplied by the emergency power supplies. This is to prevent inacvertent coren dilution during certain Occurrences in which electrical loads are disconnectec from, and returned to, the Emergency Buses. Provision should be mace so that at the end of refueling, before start-up, a surveillance procedure will confirm

.that this Licensing Basis FSAR requirement continues to be met. Absence of confirmation of this LCO is a non-conservative condition; the licensee sna11 evaluate and propose. '

T.S. Item 3/4 9.8 RESIDUAL HEAT REMOVAL AND COOLANT CIRCULATION; HIGH WATER LDEL -

The LCO provides that:

3.9.8.1 At least one residual heat removal (RHR) loop shall be OPERABLE anc in operation.*

The Licensing Basis, reference 20, Page 5.5-23, uncer Refueling, and

' page 5.5-24 under 5.5.7.3.1, System Availaoility anc Reliacility, last caragrs:n, shows the ifcensing of the RHR systam is never based on only one RHR systaa being ooersele. Two are always to be available. This preposal is therefore outsice the LCO for the FSAR in a non-conservative manner. The Licensee snali evaluate and propose .-

In his Basis, on T.S. Page 3/4 9-2, last para. , the if censee has preposed that: g "With the reactor vessel head removed and 23 feet of water above the reactor vessel flange, a large heat sinx is available for core cooling.

Thus, in the event of a failure of the operating RHR loop, aceouate time is provided to initiate emergency procedures to c:ol the core."

In the FSAR, reference 8, page Q 212-56 under Case 2, it has been estimated that on loss of all RHR Cooling due to a fail closed RHR/RCS isolation valve, it will take 2's hours for the available water inventory to boil. In that case, a numcer of alternates are proposed to resolve the situation and almost invariably, electric power is required, and in most cases the RHR equicment is used. If the basis for the licensee's request here is to encole him to coerste -

06/01/84 105 Revision A

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with only one available electrical bus, it is unacceptable, as the loss of one operable RHR on loss of the only available electrical bus, with containment "

isolation required .in 2h hours, has not been evaluated. At this time we have l

no acceptsele safety basis for allcwing the proposed deviation from the Limiting l

Conditions of Operation of the Licensing Basis FSAR wnich is tnat 2 RHR loops from separate emergency buses be operable. The proposal is therefore non-conservative and the licensee must evaluate and propose.

I Furthermore, the licensee must provide that the level of water in or above the a loops be such as to provide acceptable flow, including NPSH conditions, at inlet to the RHR pumps. Absent those required conditions from the Limiting -

Conditions of Operation could make them non-conservative. The ifcensee shall evaluate and propose. .

f f The ACTION STATEMENT provides that with no RHR loop operable, the containment L should be closed within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />. Information in referenca.S. page Q 212-56 under Case 2 shows that if RhR is aDsent [by isolation of the RCS/RHR inlet g valve] that:

i "Approximately 2.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> ire available to the operator to establish an alternate means of core cooling. This is the time it would take to heat i 300,000 gallons of water in the refueling canal from 140*F to 212*F, assuming the maximum 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> decay heat load." ' .

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The current value of 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> appears less conservative than this calculated value of 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> within the FSAR. The licensee shall evaluate and propose. 9 The current surveillance requirement:

4.9.8.1 "At least one RHR loop shall be verified to be in operation and circulating reactor coolant at a flow rate of greater than or equal to 3000 gpm at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />."

is ceficient in that tne thermal pe! Mrmance of any one RHR system to Licensing Basis safety recuirements is not being verified. The T.S. is therefore non-c:nservative with resoect to the Licensing Basis. The licensee shall evaluate and propose.

i Footnote *: The licensee also proposes that. -

"The [only operable] RHR loop may be removed from operation for uc to 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> per 8-hour period during the perfonnance of CORE ALTERAT!CNS in the vicinity of the reactor vessel hot legs."

The licensee shall provide the basis for this proposal including safety evaluation, any related compensating actions, and a related proposal. [It should be noticed that such an action could increase pool temperature by 35' i

and in so doing decrease the available response to handle a loss of cooling

! cacacity from 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> down to lh hours, and for a considerable period of time l thereafter whilst temperatures are again being reduced to tne required value f

' of 140*F.] This proposed T.S. is outside the Licensing Basis in a nonconserva-tive manner. The Licensee shall evaluate and propose.

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106 Revision A 06/01/84

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Review of available responses to the consecuences of a fail closed KCR/RHR isolation valve, include many procedures using the containment sump. To allow for this single failure contingency, the licensee should therefore ensure tnat the containment sump will be operable during this mode, and with an appropriate I surveillance procedure. There should also be provision for available fire pumps and necessary hoses to be assuredly available to enable use of the alternate procedures which have been described in reference 8, pages Q 212-56 and 57, revision 25. The current T.S. must be considered non-conservative.

licensee shall evaluate and propose.

T/S Pace 3/4 9-12 REFUELING OPERATIONS r .

The subtitle should read as 3/4.9.9 HIGH WATER LEVEL Clarify.by adcition of the term HIGH T/5' Pace 3/4 9-11 REFUEL'ING OP'ERATIONS L0W WATFR LEVEL APPLICABILITY: MCOE 6 when the water level above the top of the reactor vessel flange is less than 23 feet.

GENERAL REVIEW: Whereas the existing FSAR under reference 20, page 5.1-7 discusses Refueling, it does not provide for a sustainec period of normal operations under these Low Water Level conditions. The FSAR provides that:

"Refuelinc Before removing the reactor vessel head for, refueling, the system temperature has been reduced to 140*F or less and hydrogen and fission product levels have been reduced. The Reactor Coolant System is then drained until the water level is balow the reac;2r vessel flange. The vessel head is then esised as the refueling canal is flooded. Uoon '

completion of refueling, the system is refilled for startup." ,

Furthermore, we find that the FSAR analyses of the single failure of the RHR/RCS isolation valve is not predicated upon operations at " Low Water Level" so that no specific analyses and/or protective actions have not been deveicoec for these circumstances. However analyses have been undertaken for the water inventories and temperatures in the RCS system that might apply under those conditions. Presumacly therefore, the "0PERATING MODE - LOW LEVEL" is a long term changing condition following Cold Shutdown, with locos drained and bolts tensioned changing to bolts untensioned acd removal of the head, as concomitant flooding of the reactor vessel cavity continues. At this time therefore, we cannot presume that the consequences of the case of single failure of the RHR/RCS isolation valve used as Case 3 in FSAR reference 8, page Q21-57, does not also apply under this MODE. We will use these consequences to evaluate.

Further, since this is affectively a long term changing condition, in the FSAR, it is not accepta01e to allow scme of the provisions requested such as one nour for the performance'of CORE ALTERATIONS- which by T.S 3/4 9.9 are only permissible under tnat specification with at least 23 feet of water over the reactor vessel flange.

I 06/01/84 107 Revision A

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It is proposed that an additional item be added to the current statement of APPLICABILITY to the effect that: This MODE shall not to be used for continuous .,

normal operations, but only as a set of circumstances occurring during the I

period in which the Reactor Vessel Head is being untensioned and removed and the reactor cavity and refueling canal are being filled, and the same volumes 1 are being drained for replacement and tensioning of the Reactor Vessel Head.

The licensee sna11 evaluate and proposa.

E The existing LCD specifies that:

"3.9.8.2 Two independent residual heat reeval (RHR) loops shall be ,_

OPERABLE, and at least one RHR loop shall be in operation.*" .

Additionally, the current FSAR requires that each of the RHR trains be provided .

with power from two (2) redundant electrical buses so that each pump receives . Without i power from a different sourca; re.fe.rence 20, page 5.5-24, revision 9. -

this requirement, the T.S is less conservative than the F5AR and the licensee * ~

[

' 'shall evaluate and propose.'

Additionally, the current FSAR, reference 8, page Q212-57, revision 25, describes ,

that in the event of loss of flow caused by closure of the RHR/RCS isolation  !

valve, [and also by cessation of flow in the system]

"The operator would.be alerted to the loss of RHR ficw by the RHR a low flow alarm.

Assuming worst case conditions (maximum 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> decay heat,"--and the RCS drained to just below the vessel flange) and making conservative assumptions about the amount of water available to heat up and boil of f, if the operator took no action, boiling would begin in about five minutes, the water level in the vessel would be down to the level of fuel in about 100 minutes."

In the event only 1 RHR loco is recuired to be in^ operation, the LCD should therefore recuire 2 operacle safety related RHR low flew alarms on each single operating system so that the operatorIscan thisrescend withinexcessive time frame 10 minutes to commence since boiling operation of the redundant system.It is necessary to maintain two operating RHR systems so will have commenced. The licensee shall evaluate that boiling will not occur with a single failure.'

ss and propose.

Aceitionally, the above information defines an LCO of a minimum A further requirement (LCO) is that any such minimum volume should te such that including the level of water in or above the loop provides acceptable flow, i

NPSH conditions, over the range of temperatures expected at inlet to the RHR pumos. Absent those required conditions from the Limiting Conditions The of Opera-tion makes them non-conservative in respect of the Licensing Basis.

j licensee shall evaluate and propose.

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Revision A 108 06/01/84 -

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s Footnote *: providas.that,

"* Prior to initial criticality the RHR icop may be removed from cpera-tion for up to 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> per 8-hour period curing the performance of C RE ALTERATIONS in the vicinity of the reactor vessel hot legs."

This is an invalid request as all CORE ALTERATIONS are only permissible under TS 3/4 9.9 HIGH WATER LEVEL - REACTOR VESSEL. This is a ncn-conservative T.S proposal. The Licensee shall propose and evaluate.

Item 4.9.8.2, a surveillance requirement, specifies:

"At least one RHR loop shall be verified in operation and circulating reactor coolant at a flow rate of greater than or equal to 3000 gpm at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.". -

A time delay of 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> is excessive to verify a loop in operation, and this has been considered earlier in this section.

Further, the surveillance requirement, every 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />, is intended to ensure not only that the system is operating, but that it is operating at process conditions, including instrumentation and control, which can be evaluated to show that the equipment is capable of performing its Licensing Basis safety function. The current recuirements for this item are aesent most of this information; it is therefore non-conservative and the licensee shall evaluate and pronose.

T r, The current ACTION STATEMENT calls for containment closure in 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> [i.e.

240 mins). Earlier conservative calculations for this MODE show that loss of all RHR in this MODE can cause boiling in 5 mir:utes and core uncovery in 100 mins. Given the circumstances, containment enclosure should be effected immediately, conmencing RHR low flow alarms. The licensee shall evaluate, anc propose. The current T.S. accears nonconservative with resoect to't.5e Licensirg 3 asis.

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06/01/84 109 Revision 2

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Addenda e

T.S. SECTICN 3/4.5 EMERGENCY CORE COOLING SYSTEMS T.S. SECTION 3/4.4.4.1 RCS LCOPS AND CCOLANT CIRCULtTICN/ HOT SHUTCCWN MCOE 4 More recant information, and a detailed check on certain elements of the proposed T.S. relevant to the above section, and the Licensing Sasis FSAR, E and particularly referenca 5, Section 7.4.1.6 Emergency Core cooling Systems and Section 7.4.1.5 Residual Heat Removal System, does not appear to provide acceptable surety that:

a) The Reactor Coolant Pressure Boundary (RCPS) valves on the RHR/RCS suction line are confirmed closed in MCDES 1, 2, & 3.

b) That the RC.PB valves in the RHR/RCS suction line are' individually identified as opened in the RHR MCCE. l t

I c) That in RHR MODE 4,the RHR system must be caoacle of automatic l i

re-alignment to the ECCS mode with residual ECCS equipment, in the I event of a SI signal, including automatic closure of the RCPS Isola-tion valves on the RHR/RCS Suction Line in accordance with 10 CFR 50 i Acp A Criterion 55(4) and subsequent automatic opening of valves to the [

J RWST in accordar.ca with 10 CFR 50 App A, Criterion 20 [with appro-priate provision for RHR pump protection].

The current position in respect of c above appears to be absent those

• requirements and therefore non-censervative. The Licensee shall evaluate and propose.

The T.S. should provide the LCOs and surveillance in the overpressurization j I

! protection system of the RHR system as described in Licensing Basis FSAR, reference 3, page 5-5-24.

it provices Procosed T/S Page 3/4 5-5, item 4.5.2.d, 1) b) accears incorrect:

that, in establishing ECCS operacility:

d. At least once per 18 months by:

1) Verifying automatic isolation and interlock action of the RHR .

System from the Reactor Coolant System by ensuring that:

a) With a simulated or actual Reactor Coolant System pressure signal greater than or equal to 425 psig the intarlocks prevent the valves from being openeo, and b) With a simulated or actual Reactor Coolant System pressure signal less than or equal to 560 psig the interlocks will cause the valves to automatically close.

Item b) acove is incorrect in that it should ensure that with a simulated or actual Reactor Coolant System pressure signal greater than 475 psig, tne 110 Revision A 06/01/84

interlocks will cause the valves to automatically close, reference 4, section 5.5.7.3 ' and reference 5, section 7.4.1.5.4. dVAs Towh non -Coor:&rd. rG.

,4/Je;/he pr posed T.S. closes the valves wnen they are in fact required to be coen and is/ therefore non-conservative. Further, che lower pressure of agg/nj 475psigrequiredtocloseismoreconservativethanava)/ eof 560unless there are set Point and Channel considerations - The pressure is less conser-vative than the Licensing Basis FSAR value.

06/01/34 111 Revision A

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LIST OF REFERENCES

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1. Letter from H. B. Tucker (D.P.Co) to H. R. Denton (NRC) dated September 27, 1982 to the subject of "McGuire Nuclear Station."

2. Memo from C. O. Thomas (SSPS) to Brian W. Sheren (RSB) on the subject of

" Proof and Review of McGuire - Units 1 and 2, Technical Specifications."

-Dated January 14, 1983. B

3. U.S. Nuclear Regulatory Commission, Final Safety Analysis Report, Volume 4, Duke Power Company, McGuire Nuclear Station, Units 1 and 2.

4 U.S. Nuclear Regulatory Commission, Final Safety Analysis. Report, Volume 5, Duke Power Company, McGuire Nuclear Station, Units 1 and 2, Rev. 45. .

5. U.S.. Nuclear Regulatory Commission, Final Safety Analysis Report, Volume 7, ,

Duke Power Company, McGuire Nuc-lear Station, Units 1 and 2, Rev. 45. l

6. U.S. Nuclear Regulatory Commission, Final Safety Analysis Report, Volume 8, Duke Power Company, McGuire Nuclear Station, Units 1 and 2, Rev. 45.

7. U.S. Nuclear Regulatory Commission, Final Safety Analysis Report, Volume 10,  !

Duke Power Company, McGuire Nuclear Station, Units 1 and 2, Rev. 45.

8. U.S. Nuclear Regulatory Commission, Final Safety Analysis Report, Volume 11, Duke Power Company, McGuire Nuclear Station, Units 1 and 2, Rev. 45.

9. Deleted

10. U.S. Nuclear Regulatory Commission; Office of Nuclear Reactor Regulation;

" Safety Evaluation Report; McGuire Nuclear Station Units 1 and 2, Duke Power Company," NUREG-0422, on Occket Nos. 50-369 and 50-370, March 1, 1978.

11. U.S. Nuclear Regulatory Commission, Office of Nuclear Reactor Regulation,

" Safety Evaluation Report, McGuire Nuclear Station Units 1 and 2, Duke Power Company," NUREG-0422, Supp. 1, on Occket Nos. 50-369 and 50-370, May 1978. -

12. U.S. Nuclear Regulatory Commission, Office of Nuclear Reactor Regulation,

" Safety Evaluation Report, McGuire Nuclear Station, Units 1 and 2, Duke Power Company," NUREG-0422, Sucp. No. 2, on Oceket Nos. 50-369 and 50-370, March 1979.

13. U.S. Nuclear Regulatory Commission, Office of Nuclear Reactor Regulation,

" Safety Evaluation Report, McGuire Nuclear Station, Units 1 and 2, Duke Power Compas ,," NUREG-0422, Supp. No. 3, on Occket Nos. 50-369 and 50-370, May 1980.

14 U.S. Nuclear Regulatory Commission, Office of Nuclear Reactor Regulation,

" Safety Evaluation Report, McGuire Nuclear Station, Units 1 and 2. Duke Power Company," NUREG-0422, Supp. No. 4, on Occket Nos. 50-359 and 50-370, January 1981.

Revision A 112 06/01/84

15. U.S. Nuclear Regulatory Commission, Office of Nuclear Reactor Regulation,

" Safety Evaluation Report, McGuire Nuclear Station Units 1 and 2, Duke Power Company," NUREG-0422, Supp. No. 5, on Docket Nos. 50-369 and 50-370, April 1981.

16. Memo from R. W. Houston to T. M. Novak on the subject of " Staff Review and Input to SER Supplement No. 6 for McGuire Nuclear Station Units 1 and 2". Dated February 08, 1983. ( d" : k - 7 / )

17. Letter from H. B. Tucker (D.P.Co) to H. R. Denton (NRC) on the subject of McGuire Nuclear Station, Units 1 and 2, filing amendment No. 71 to its Application for License for the McGuire Nuclear Station and Submitting

, Revision 45 to the Final' Safety Analysis Report. Dated February 16, 1983.

18. Letter from W. O. Parker (0.P.Co) to H. R. Denton (NRC), dated Oct. 3, 1981 on the sucject of McGuire Nuc, lear Station, Unit 1 and submitting copies of Report identified as " Westinghouse Reactor Protection System /

Engineered Safety Features Actuation System Setpoint Methodology, Duke Power Company, McGuire Unit 1," by C. R. Tuley et al. and datec April 1981, published by Westinghouse Electric, Nuclear Energy Systems, PROPRIETARY.

19. Westinghouse Electric Corporation, PWR Systems Division " Westinghouse Emergency Core Cooling Sy, stem - Plant sensitivity studies, WCAP-8356.

August 1,1974.

20. U.S.NuclearRegulatoryCommission,FinalSafetyAnalysisReport, Volume 5, Duke Power Company, McGuire Nuclear Station, Units 1 and 2, Rev. 45.

21. Letter from T. M. Novak (NRC) to H. B. Tucker (D.P.Co), dated May 17, 1983 on the subject of OL Condition 2.C.(11)g, Anticipatory Reactor Trip (II.K.3.10) (McGuire Nuclear Station, Unit 1).

22. U.S. Nuclear Regulatory Commission, Final Safety Analysis Recort, Volume 9.

Duke Power Company, McGuire Nuclear Station, Units 1 and 2, Rev. 45.

23. Letter from W. O. Parker (0.P.Co) to H. R. Denton (NRC), datec August 13, 1980, re: McGuire Nuclear Station.

24. Letter from W. O. Parker (D.P.Co) to H. R. Denton (NRC), dated Septamcar 13,

. 1980, re: McGuire Nuclear Station. Page 13 Response to 3(e).

25. Duke Power comoany McGuire Nuclear Station, Unit 1. Docket Nc. 50-369, License No. NPF-9 Startup Report, February 15, 1982.

26. Memo for RSB, CPB, ICSB Members from Brian W. Sheron (RSB), Cari H.

Berlinger (CPB), Faust Ross (ICSB) dated April 12, 1983 on the Suoject 1-of Inadvertent Boron Dilution Events. ( 03*' ' 2)

27. Westinghouse Electric Corporation, Nuclear Energy Systems Topical Report, Overpressure Protection for Westinghouse Pressurized Water Reactors, WCAP-7769, Rev. 1, June 1972.

06/01/84 113 Revision A

• O pjk7 1  % :* f F- I ,

!- M1 y Ay

+p on

28. AWestinghouse Electric Corporation for the Westinghouse Owners Group #Y-Reactor Coolant System Overpressurization, July 1977. PROP Alf T4 I"

29. U.S. Nuclear Regulatory Commission, Final Safety Analysis Report, Volume 6, Quke Pcwer Company, McGuire Nuclear Station, Units 1 and 2, Rev. 45.

E l

i.

T 1

t 1

l l

l l

l 114 Revision A 06/01/S4 i

(

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7 TABLE 1 SECTICNS REVIEWED BY REACTOR SYSTEMS BRANCH SECTION PAGE 2.1 SAFETY LIMITS 2.1.1 REACTOR CORE ................................................... 2-1 2.1.2 REACTOR COOLANT SYSTEM PRESSURE ................................ 2-1 FIGURE 2.1-1 REACTOR CORE SAFETY LIMIT - FOUR LOOPS IN OPERATION . . .. . 2-2

2. 2 LIMITING SAFETY SYST M SETTINGS 2.2.1 REACTOR TRIP SYSTEM INSTRUMENTATION SETPOINTS ............. .... 2-4. .

TABLE 2.2-1 REACTOR TRIP SYSTEM INSTRUMENTATICN TRIP SETPOINTS .. .... 2-5 3/4.0 A P P L I CA B I L ITY . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 /4 0- 1 3.4.1 REACTIVITY CONTROL SYSTEMS 3/4.1.1 BORATION CONTROL -

Shutdown Margin - T,yg > Programmed No Lead T,yg ........... 3/4 1-1 , ,

Shutdown Margin - T < Programmed No Load T and > 20 0* F . . ay g , , , , , , , , , , , , , , , , , , , , , , , ay g, , , , , , , , , , , , ,

Shutdown Margin - T avg < 200*F ......................... ... 3/4 1-3 Moderate Temperature Ccefficient ............................ 3/4 1-4 Mi nimum Temperature for Cri ticality . . . . . . . . . . . . . . . . . . . . . .. 3/4 1-6 3/4.1.2 SORATION SYSTEMS Flow Path - Standbye, Shutdown anc Refueling ............,... 3/4 1-7 Flow Paths - Power Operation, Startup, Stancbye down to 1000 psig/425' F ................................. .. . 3/4 1-3 Charging Pump - Standbye, Shutdown and Refueling ...... .... 3/4 1-9 Charging Pumps - Operating ................................. 3/4 1-10 Borated Water Sources - Shutdown ........................... 3/4 1-11 Borated Water Sources - Ocerating ............. ............ 3/4 1-12 Instrumentation .............. .... ......... ... .. ..... 3/4 1-13a 06/01/84 115 Revision A

. . ~ ~ - . ... . . ~ . . . . -- . . . - - - .- .

PAGE SECTION TABLE 3.1-1 ACCIDENT ANALYSES REQUIRING REEVALUATION IN THE EVENT 3/4 1-16 OF AN INOPERAB LE FULL-LENGTH ROD . . . . . . . . . . . . . . . . . . . . . .

Position Indication Systems - Operating .................... 3/4 1-17 Position Indication System - Shutdown ...................... 3/4 1-13 a

. Rod Drop Time (Units 1 and 2) .............................. 3/4 1 ,19 3/4 1-20 Shutdown Rod Insertion Limit (MODES 1 & 2) ................. .

Shudown Rod Insertion Limits (Modes 3 - 5) . . . . . . . . . . . . . . . . .

Control Rod Insertion Limits ......................,.........

3/4 1-21 ,

l 3/4.2 POWER DIS'T9IEUTION L'IMITS t i

3/4 2-16 l TABLE 3.2-1 DNB AND REACTOR COOLANT SYSTEM PRESSURE PARAMETERS . . .. .

3/4.3 ' INSTRUMENTATION 3/4 3-1 3/4.3.1 REACTOR TRIP SYSTEM INSTRUMENTATION .........................

3/4 3-2 l TABLE 3.3-1 REACTOR TRIP SYSTEM INSTRUMENTATION .....................

3/4 3-9 l TABLE 3.3-2 REACTOR TRIP SYSTEM INSTRUMENTATION RESPONSE TIMES ......

I TABLE 4.3-1 REACTOR TRIP SYSTEM INSTRUMENTATION SURVEILLANCE 3/4 3-l' REQUIREMENTS ..........................................

3/4.3.2 ENGINEERING SAFETY FEATURES ACTUATION SYSTEM 3/4 3-15 INSTRUMENTAT!CN ..................................... .....

TABLE 3.3-3 ENGINEERED SAFETY FEATURES ACTUATICN SYSTEM .... .. 3/4 3-16 INSTRUMENTATICN ..............................

TABLE 3.3-4 ENGINEERED SAFETY FEATURES ACTUATION SYSTEM .... 3/4 3-25 INSTRUMENTATION TRIP SETPOINTS ..................

3/4 3-30 TABLE 3.3-5 ENGINEERED SAFETY FEATURES RESPONSE TIMES . . . . . . . . . . . . . . .

3/4.4 REACTOR CCOLANT SYSTEM 3.4.4.1 REACTOR COOLANT LOOPS AND COOLANT CIRCULATION 3/4 4-1 Startup and Power Operation ................................

3/4 4-2 Hot Standby ................................................

3/4 4-3 Hot Shutdown .................. ............................

116 Revision A 06/01/84

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SECTION PAGE Cold Shutdown - Lceps Filled ...................... ........ 3/4 4-5 Cold Shutcown - Locps Not Filled ....................... . . 3/4 4-6 3/4.4.2 SAFETY VALVES

. Shutdown ................................................... 3/4 4-7 Operating .................................................. 3/4 4-8 3/4.4.3 PRESSURIZER ................................................ 3/4 4-9 3/4.4.4 R E L I EF VA LV E S . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3/4 4-10 3.4.4.5 STEAM GENERATORS ........................................... 3/4 4-11 Pressurizer ................................................ 3/4 4-35 Ove rpres sure Protecti on Systems . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3/4 4-36 3/4.5 EMERGENCY CCRE CCOLING SYSTEMS 3/4.5.1 ACCUMULATORS Cold Leg Injection ......................................... 3/4 5-1 Upper Head Injection ....................................... 3/4 5-3 l

l 3/4.5.2 ECCS SUBSYSTEM - T,yg > 350*F .............. .............. 3/4 5-5 3/4.5.3 ECCS SUBSYSTEMS - T,yg 5 350*F . ........................ .. 3/4 5-9 3/4.5.4 BCRCN IN'ECTICN TANK (Unit 1 Only) ...................... . 3/4 5-11 3/4.5.5 REFUELING WATER STORAGE TANK . . . . . . . ............ ..... . 3/4 5-12 3/ 7 PLANT SYSTEMS 3/4.7.1 TURBINE CYCLE Safety Valves Tureine Trip on Reactor Trio ............ ... 3/4 7-1 Auxiliary Feedwater System ..... ........................... 3/4 7-4 Auxiliary Feedwater Condensate Storage System ... ....... 3/4 7-5(a)

Main Steam Line Isolation Val ves . . . . . . . . . . . . . . . . . . . . . . . . . . 3/4 7-S Atmospheric Dumo Valve .................................... 3/4 7-Ba 3/4.7.2 STEAM GENATOR PRESSURE / TEMPERATURE LIMITATION . .... ....... 3/4 7-9 06/01/84 117 Revision A

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PAGE SECTION 3/4 7-10 3/4.7.3 COMPONENT CCOLING WATER SYSTEM ............................. = l 3/4 7-11 3/4.7.4 NUCLEAR SERVICE WATER SYSTEM ...............................

3/4 7-12 3/4.7.5 STANDBY NUCLEAR SERVICE WATER POND .........................

3/4.9 REFUELING OPERATICNS 3/4 9-1 3/4.9.1 BORON CONCENTRATION ........................................

3.4.9.2 INSTRUMENTATION ............................................ 3/4 9-2 P 3/4.9.8 RESIDUAL HEAT REMOVAL AND CCOLANT CIRCULATION

• High Water Level .................................."......... 3/4 9-10 3/4 9-11 Low Water Level ...........................................

113 Revision A 06/01/84 1

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. r TABLE 2 TECHNICAL SPECIFICATION PAGES AFFECTED The following pages of the Technic?1 Specifications are affected by this review:

T.S. Pages 2-1, 2

TABLE 2.2-1, T.S. Pages 2-5 2-6 2-7 T.S. Pages 3/4 1-1 3/4 1-2 3/4 1-2a proposed ,

3/4 1-6 3/4 1-7 3/4 1-8 3/4 1-9 3/4 1-10 3/4 1-11 3/4 1-12 -

3/4 1-13 3/4 1-13a) 3/4 1-20a) ,

3/4 1-21 T.S. Pages 3/4 2-15 16 TABLE 3.3-1, T.S. Pages 3/4 3-2 3-3 3-4 3-5 3-6 TABLE 3.3-2, T.S. Pages 3/4 3-9 -

3-10 TABLE 3.3-3, T.S. Pages 3/4 3-16 3-17 3-18 3-19 3-20 3-21 3-22 3-23 06/01/84 llo Revision A i

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TABLE 3.3-4. T.S. Pages 3/4 3-25 3-26 .,

3-27 3-28 3-29 TABLE 3.3-5, T.S. Pages 3/4 3-30 3-31 E 3-32 3-33 T.S. Pages 3/4 4-1 .

4-2 '  ;~

4-3 i

~

4-4 -

4-5 *

. 46 4-6(a) proposed 4-7 4-8 4-9 4-10 l 4-11  ;

5 4-36  :

3/4 5-1, T.S. Pages l 5-2 5-2a) proposed  ;

5-2b) proposed 5-3 5-4 5-da) proposed 5-4b) proposed 5-5 5-6 -

5-S 5-9 5-10 5-11 5-12 f T.S. Pages 3/4 7-4 7-5(a) proposed 7-5(c) proposed 7-8 7-8(a) proposed 7-10 7-11 7-12 T.S. Pages 3/4 9~1 9-10 9-11 -

9-12 Revision A 120 06/01/84 ,,

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, e APPENDIX A t

TECHNICAL SPECIFICATIONS SELECTED RELEVANT REGULATIONS y I 30'.11 Title 10 anergy mined that there are no unresolved (1)(1) The processing fabricadon or safety issues relaung to the addluonal refining of special nuclear matertal or activities that may be authortsed pur. the separauon of special nuclear mate.

suant to this paragraph that would rial. or the separacon of special nucle.

consutute good cause for withholding at material from other substances by a authertsauen. prune contractor of the Department (4) Any acuvtues undertaken pursu. under a prime contract tor:

ant to an authchon gr2nted uder (A) The performance of work for the this paragraph ahan be entirely risk of the *pplicant and except as to

&& the Department at a Umtad States govern.

ment-owned or controlled atte; matters determined un,, der paragraphs (3) Research in. or development.

[.

r' de'no manufactura, storage. testing or trana, a

the issuance of a construction pernut portauon of. stomic weapons or com.

with respect to the requirements of ponents . hereof; or .

the Act, and rules, regulacons, or (C) The use or operation of a pro.

orders promulgated pursuant thereto. duction or uulissuon facility in a (Soun.101.188. SS Stak 938. 988, as amended (42 UAC. 2131. 2238: see.102. Fue.1. St.

or /

19e. 83 StaL 883 (42 U.S.C. 43323: see. 2o1. as (!!) By a prime contractor or subcon.

amendee. Fut. L 93-438. 88 Stat 1242. Fug, tractor of the Comaussion or the De.'

I. 94-79. 99 Stak 413 (42 CAC. 38413: sec. partment under a prtme contract or 161 as amenese. Fue. L 83 7o3. 88 Stak 948 subcontract when the Ca==haa de.

(42 U.S.C. 22011) ternunes that the esempuon of the (21 FR 388. Jan.19.1964, as amended at 28 prime contractor or subcontractor is FR 3712. Sept. 9.1980: 32 yR 2381. Jan. 31. authertsed by law; and that, under the 8*

.2 9 asE 808. . 24. s1 3 terms of the contract or subcontract.

there is adequate assurance that the FR 282*9. July 18.1974: 39 FR 33202. Sept.

18.1974: 42 FR 22881. 3Rar 8.1977: 43 FR work thereunder can be asemymahad 8024. Fee.17.19183 without undue risk to the public health and safety; 4 38.11 Eseections and esemptions from (2)(1) The construccon or operauen Mag rguments, of a produccon or utilisation facility Nothing in 12us part shall be deemed for the Department at a United States to require a license for: government-owned or controlled site.

(a) The manufacture, producuan, or inclualag the transportation of the accuis:ticn by the Oopartment of Oe. produccon or uuhzation factlity to er fense of any unli:stion facil!!y author. from such site and the performance of 12ec pursuant to secuon 31 of the Act, contract services during temporary in.

or the use cf such fact!!!y by the Oe. ter;.:stiens of such transportauon; or partment of Defense or by a persca the construction or operacon of a pro.

under contract with and for the ac. ducten or ut:11 stien facility for the count of the Department of Oefense: Department in the performance of re. p (b) Except to the extent that Aamun* search In. or development, manufac.

1strauon facilities of the types sueject ture. storage, testing. or transporta.

• to licensing pursuant to section 202 of tion of. Stonue weapons or components the Energy Reorgamasuon Act of thereof: of the use or operacon of a 1974 8 are mvolved: produccon or unlization facility for the Department in a United States

'The Department fact!!ues teenuflee in government. owned vehicle or vessel:

Ni enstrsuon Iaeuse Metal Fast ded. T at such act1Mes are cm.

Breeeer reactors unen ooerstee as cart of ducted by a prime contractor of the the power recersuon fae:J:ues of an electrie uutay statem. or unen operssee as any 1973. wnen opersted as part of the power other manner for tne purpose of eemon. genersuon fact 11ues of an eteetne uu!ty stra::ns tne sunatcty for cornmercial ao. system. or when operates :n any otner 311 canon of ruen a reactor. manner for tne purpose of demonstrat:ns (2) Other comonstrauen nuclear reacters, the suntedity for commere:aa a;stictuen of except those Ln emstence on January 19. sucn a reactor.

392 -

06/01/84 121 Revision A

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Cispter 1-.Nudear 2egulatory Commission } 50.21 Oepartment under a prime contract meet delay those needs costs on aapplicant to the timely hasis andand to witn the Department.

(u) The contrucuan or operacon of ccnnmers.

a production or ut31sauon faccity by a Issuance of such an exemption shan 3 petme contractor or subcontractor of not be deemd to consutute a commut.

the Commission or the. Departmtat mnt to !ame a constmetion permat.

under his prime contract or subcon. During the period of any exemption tract when the Commission deter. granted pursuant to this paragraph p-mines that the exemption of the .(b). any activities conducted shall be r prime contractor or subcontrator is carried out in such a manner as w1U author *. Zed by law; and that, under the mimmmm or reduce their environmen. t terms of the contract or subcontract, talimpact.

there is adequate assurance *.ast the- .

t Tcrk thertunder can tie accoinctished (37 y!L $t43. Mar. 21.1972. as ame'nded at without uncue risat to the public :s 75t :s:n. Jmt is.1974. 40 ya 3 ss. :. tar. '

health and safety.- 3.ISTS)

(c) The transportsuon or possesaton 3 54.13 Anacks and dentrective acts br en.

of any produccon or utussauon factu. emies of the Umtad States: and defense ty by a common or contract carrier or arsm Jes, g

warehousemen in the regular course of carriage for another or storage inc1* An applicant for a license to ecc.

dent thereto. struct and operate a production or uu. g lisation faculty, or for an amendment (to TR 8T84. Mar. 3.1975) to such Heense,is not required to pro. f 8 54.12 .Spesille esempsiene. vide for rian2fn features or other mesa.

ures for the spec:f!c purpose of protec. l (a) The Comnussion may, upon sp.i uon sgstr.at the effects of (a) auacas  ;

pucation by any interested person or and destructive acts. including sabo.

upon its own in1Mative. grant such ex. tase, directed against the facility by empuona from the requartments of an enemy of the United States,wheth.

the regulauons in this part as it deter. er a foreign government or other

' mined are authorised by law and wd1 person, or (b) use or deployment of not, *= danger life or property or the weapons incident to U.S. defense acuv.

common defense and secur:ty and~are Otherwise in the pubile interest. , itica.

(b) Any person may request an ex. (22 FR 13448. Sept. 28.19sT1 emption per itting the conduct of ac.

tavlues pr".or to tne 'Sauance

. of a con

• C*.AasIr! CAT!o,N AND DEscS!rT oM cr ,

struc* en permit pronabated by 150.10. ,;=g33 3 The Com-tmon =ay grant suca an -

exempuon upon consider:ng and bal* 33023 Twe efsense of ileensee.

anc:nz the f:u:w.ng f actors:

censes d be issued to na=ed >

I l

(1) Whether conduct of the proposed sces applying to the Comm.ssion

' acuvities wul give r:.se to a s:gn:ficant thertier, and wiu be either class 104 cr adverse im:act on the environment. claas 103.

and the nature and extent of such ,

M h of any adva iSoll Case 164 !!eenees: for

'A'? ""d "'"" **d medical d'**"'

environment imbact from conduct of #*U

the proposed activities can reneonably

- A class 104 ucesse wul be tasued. to be effected should such redreas be noe. an appucant who quallfles for any one ensary:

(3) Whether conduct of the proposed or more of the followtas: to transfer or activttles would foreciose subsequent receive in interstate commerce. manu.

facture, produce. transfer, acquire.

adoption of alternauves; and possess, or use, (4) The effect of delay in conducting such activtues on the pubile interest. medical therapy; or facil!!y for use in (a) A utuizacon including the power needs to be used (bX11 A production or usu1=auon fa.

by the proposed facdity, the svadabil. cility the construction or operation of ity of alternative sources. Lf any. to 393 f

1 122 Revision A 06/01/84

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t:

i k

(4) The information desertbed in minimum information

• to be included I paragraphs (aX1) and (2) of this sec. shall consist of the foUowing: I s

' uon shan be submitted as a separate (1) A descripuon and safety assess.

document prior to any other part of ment of the site on which the facality the license af9asma*'aa as provided in is to be located, with appropriate at.

paragraph th) and la assordance with tantion to features affecting faculty

, i 2.101 of this shapter. design. Spesial attsnuon abould be di.

(b) Essent as provided in paragraph rooted to the site evalumuon facton (d), any person who appiles for a class identtfled in Part 100 of this chapter.

103 construccon perrait for a nuclear such amman== mar shall contaan an anal.

power reactor on or after July 2s, it'rs ysas and evaluauen of the major strue.

shan subaut the document uued "In* tures systems and components of the '

formatton Requested by the Attorney General for Anutrust Review" at least facility wtuch bear significantly on the *

. nme (9) months but not more than acceptability of the sate under the site tharty.sta months prior to the date of evaluauon factors idenufled in Part

• sutunsttal of any part of the applica. 100 of thas chapter, assunung that the tion for a class N3 constnaction mate facdity win be operated at the ulti.

4 8"AE' power level whach is contemplat.

4 (88 I I ed by the applicant With respect to (d) Any person who applies for a operauen at the projected initial 1

,class 103 construction peruut for a nu. power level, the applicant is required clear power Mactor snanuant to the to sutuntt mformauen gebed in

, provtstens of i 2.101(b!) and Subpart paragraphs (aM23 through (8) of this P of Part 2 of this chapter shau section, as we!! as the information re.

substit the document title " Informs. quimd by this paragraph. In support uon Requested by the Attorney Gen. of the applicauen for a construction eral for Anutrust Review" at least permat. ,-

name (9) months but not more than (2) A sununary densr ption and dis.

thirty sis months prior to the !!!!st of cussion of the faculty, with special at.

part two or part three of the applica. tantion to design and operaung char.

Uon, whichever part is fHed first, as actenstics, unusual or novel design specalled in i 2.101(>1) of uus chap. features, and principal safety consider.

ter, acons.

(e) Any person who app!!as for 4 (3) The preliminary design of the fa.

class 103 construction pennst for a ctilty including:

uranium enricament or fuel reprocess. (1) The principal design entens for ing piant shan suomat such informa. the faculty.' Appendix A. General

' tion as may be requested by the Atter. Design Cnteria for Nuclear Power ney Genera! for anutrust review, as a Plants, estaclishes mintmum recutre.

separate document as soon as possible ments for the principal design critena and in accordance with 12.101 of this for water. cooled nuclear power plants chapter, samalar in design and locauon to planta

(See.102. Pua. L 91-190. 83 Stat. 883 (43 for which construction permits have #

j UAC. 43323: see. Sol, as amendes. Pue.1. previously been issued by the Commas.

sa-43s. as Stat.1243. Pua. L 94-79 se Stat. sion and provides guidance to appi!.

413 f 43 UAC. 344133 cents for construction permits in es.

i (39 PR 34398. Soot. 28.1974 as amended at tablishing prtactpal design criteria for 42 FR 2288T. Mar S. ISTT: 42 PR 2sT21. May other types of nuclear power unats:

19.19T1: 43 FR 4eTTS Cet. 28. Le78: 44 PR 4 setts. Cet. 22,1979)

• The neoucaat may provide intennauen 1 50.34 Consenes of appliceedeas seemassai reeuated my taas pareerson m tne term of a informe48en. siseuassen, witn seestfle references, es samt.

(as Preumisery safety analysu innues te ene earterences from. racinum et

, report. Each appucauon for -a con. pre,ieuser uiar ammaen ter =nien seeneauena nave 4

seen (Hee un the comminion.

struction permat shan include a pre. eoeneral design entena ter enemicas ummary safety analysts report. The preenamns racinues are seine oeveieoes.

j' 399 i

i 4

^

06/0U84 123 Revisien A J

..-...%.['- '#, $ ._g. _ U , b r y.%' _ e .[ [.,, -_ _,e,.w,,,-% _%,_.., ___,y_._r,

a. . ...=...: . -: == - - . : -

p Title 10-Energy

{ 50.34 (ID The design bases and the rela- the quality assurance program for a don of the design bases to the prtnet- nuclear power plant or a fuel repro-pat design criterta; cessing plant shall include a discussion (11D Informauon relative to materi- of how the applicable requirements of als of construction. seneral arrange- Appendix B will be satisfied. E ment, and approximate dimensions. (8) An idenufication of those strue.

' suffIcjent to provide reasonable assur- tures. systems. or components of the ance that the final design will conform facility,if any, which require research ,.

to the design bases with adequate and development to confirm the ade- 'i margin for safety, quacy of their design; and idenutica-(4) A preliminary analysis and evalu- tion and description of the research ation of the design ano performance and development program which will of structures, systems, and compo- be conducted to resolve any safety nents of the factitty wtth the objective questions associated with such strue.

4 of assessmg the rtset *4 public health tures, systems or components sJtd a [

and safety resulung from operauon of schedule of the research and develop- i the facility and including determins- ment program showing that such j Uon of (1) the margsna of safety during safety quesuons will be resolved at or l normal operauons and transient condi. before the latest date stated in the 40-tions anucipated during the !!fe of the plication for completion of construc- {

fact!!ty. and (11) the adequacy of strue. Hon of the factllty.

tures, systems. and componenta pro- (9) The techancal quellficauons of vided for the prevention of accidents the applicant to engage in the pro-and the mjugadon of the conse- posed activities in accordance.wlth the quences of accidents. Analysts and regulacons in this chapter. -

evaluauon of ECCS cooling perform- (10) A discussion of the applicant's ance following postulated lossof cool- preliminary plans for coping with ant accidents shall be performed in ac- emergencies. Appendix E sets forth cordance with the requirements of items which sha11 be included in these 150.48 of this part for facillues for plans, (11) On or after February S.1979 which construction permita may be lasued after December 28.1974, applicants who apply for construction (S) An idenuticadon and justifica- permits for nuclear powerplants to be built on muluunit setas shall identify tion for the selection of those varta- potent!al hazards to the structures.

bles, conditions, or other items which are determined as the result of pre- systerns and components important to Ilminary safety analysts and evatus- safety of operating nuclear facilities tion to be probable subjects of techm- from construction activttles. A discus-cal specifiestions for the factitty, with ston shall also be included of any man-special attention given to those items agertal and administrative controis which may significantly influence the that will be used during construct!on final design: Provided. Acteerer. That to assure the safety of the operaung this requirement is not applicable to unit, (b) Final safety analysts repor*. '

an appucation for a construction Each spaticauon for a lleense to oper-permit filed arter to January 18.1969. ate a factitty. shall include a final (8) A preliminary plan for the 40311 ,

cant's organtastion. training of person- saf=ty analysts report. The final safety nei, and conouct of operations. analysts report shall include informa-

47) A desertption of the quality as- tion that desertbes the facility. pre-surance program to be applied to the senta the desten bases and the !!mita design fahrtestion, construction. and on its operation, and presents a safety testtrtt of the structures. systems, and analysis of the structures systems, components of the facility. Appendix and componenta and of the facility as B. " Quality Assurance Critetta for Nu- a whole and shad include the follow- i clear Power Plants and Fuel Repro- ing:

(1) All current informadon. such as cessens ments forPlanta." sets forth the quality nasurance require- the results of enetronmental and me-programs for nuclear power planta and fuel re- teorological monitoring programs.

processing planta. The desertpuon of which has been developed since issu.

400 Revision A 124 06/01/84

%._ _ u . .. w . .. . a._ __-..m .

s e *

?. ,

Chapter 8 = Nuedeer Aa O-i Comuniesien { 50.3d ance of the construction permit relat. (3) A deserfotion and evaluation of r

ing to site evaluadon factors idenulled the results of the applicant's pro.

In Part 100 of this chapter, grams, including research and develop.

(2) A desertation and analysis of the snent. If any, to demonstrate that any structures, systems, and cogr.ponents safety questions identifled at the con.

of the facility, with emphasis upon struction permit stage have been re.

performance requirementa, the bases, solved.

' . with technical justificador therefor. (8) The following information con-upon whlen such requirerr.snts have cerning facility operation:

, been established, and the evaluations (1) The app!! cant's organizational reeutred to show that safety funetsons structure, allocations or responsibit.

will be sceompl!ahed. The description ities and suthertties, and personnel shall be suffletent to permit under. Qua11fications requirements.

standing of the system designs and (11) Manasertal and administrative +

their relationship to safety evalua.. controls to be used to assure safe oper. .

Uons. atton. Appendix B. " Quality Assurance (1) For nuclear reactors, such items Crtteria for Nuclear Power Plants and as the reactor core, reactor coolant Fuel Reprocessing Plants.* sets forth systera. Instrumentation and control the requiremnts for such controls for systems, electrical systems, contain. nuclear power plants and fuel repro.

ment system, other engineered safety eming plants. The informauon on the features, auxiliary and emergency sys. controls to be used for a nuclear power tems, power conversion systems, radio. plant or a fuel reprocessing plant shall

] active waste handling systems, and Irselude a discussion of how the appil.

• fuel handling systems shall be dis. cable requirements of Appendix B will

' be cussed insofar as they are pertinent.

(11) For facillues other than nuclear i F1 for preopersuonal testing reactors, such items as the chemical.

physical, metallurgical, of nuclear

, ,,,' ,g ,,,,,g ,,,

proem to be performed. Instrumsta. ermuona. Including maintenance, sur.

Uon arid control systems, venulauen veillance, and pertodic testing of strue.

and Ulter systems, electrical systems. tures. systems, and components. '

auxiliary and emergency systems. and (v) Plans for coping with emergen.

radioactive waste handling systems cies which sha!! Include the items shall be discussed insofar as they are specif!ed in Appendix F O'NU'"b (vi) Proposed 'acanical specif!cadons [

prepared in accorsance with the re.1 (3) The kinds and quanuties of ra*

dioactive materials expected to be pro

• eutrements of I 30.38.

duced in the operadon and the maarts (v11) Cn or after February $.1979 app!!canta who apply for operat!ng !!.

for controlling and limtung radioactive censes for nuclear powerplants to be effluents and radiation espesures operated on :nultfurut sites shall in.

Within the limits set forth in Part 20 ciude an evaluation of the potenda!

of this chapter. ~

hasards to the structures, systems. and (4) A final ncal221s ansLtzahlanc21.of components important to safety of op.

4 the design and performance of struc- ermung units resulting from construe.

tures, systeras, and components with uon activttles, as seu as a desertation the oplective stated la paragraph of the managerial and administradve (aM4) of this section N controls to be used to provide assur.

, account ahlped!Dfnt_:Diormauon de, anee that the limiting conditions for velop,st_ssnee_th_e suomattal of the pre. operauen are not eseeeded as a result 11Tigrian safety analnts mmet. Analy. of construction activities at the mul.

ses and evaluauon of ECC5 anoling Munit sites, performance following postulated loss- (7) The technical qualifications of

of. coolant aseidents shall be per. the applicant to engage in the pro.

formed in socordance witit the to. posed acuvtues in accordance with the quirements of I 50.46 for feet 11tles for regulations la this chapter.

, which a IIcense to operate may be (8) A description and plans for im.

Issued after December 28,1974. piementation of an operator rectuallfl.

401 1

06/01/84 12" Revisicn A i

l

, +.m-,w-y - ,.,-m_,- , ,e ---e-,-- -y_.4pyg, ,n, ,.,.y,m,y y ,-ww., ,qm--,,,-- -w-.,.m,,--,, - - , , - . , _ _ - - - - - , - -,,,.,g

t a

se Title 10- Inergy

} $0.34e connected to the containment atmos- tions thereof, that underlie the corre.

spending SRP acceptance criteria.

phere. (fLE4.1) ( ) ne SMP ms issg to (vti) Provide a descripuon of the management plan for design and con. criter13 f.ha&,* N "MTm_.s, tencs tat!!ish to struction activides. Lo include: (A) The use!.Irv evaluating' whether-Erra Ns appit- B organisational and management struc. cant /licenssa. -~ "e**osn r30tm.

ture sangularly responsible for direc- reger."L- - The.4RP Uon of design and construction of the tute for the regulauens . anti esetpli-

• proposed plant- (B) technical re- ance is not.a.sequesomenar.eppteenata 7 sourecs director by the appilcant* (C) shall identify differences from the

• details of the interaction of design and SRP acceptance criterts' and evaluate '

construction within the appilcant's or* how the proposed alternauves to the ganizauon and the manner by which SRP criteria provide an acceptable t the applicant will ensure close integra' . method of complytag with the Com.

• t cn of the architect engineer and the mission's regulaticns. ,

nuclear steam supply vendor: (D) pro- (Seas.1sta. letL Pub. L 83-703, se Stat.

posed procedures for handling the 3 transition to operation:(E) the degree Q . 30 g of top levet managem,ent oversteht and 5s44r. see. 7. Pue.1. s3-377. 88 Stat. 4755  ;

technical control to be exercised by sec.1911. Pua.1.83-703, es Staa, s44 tes the applicant during design and con

• U.S.C.3201n

,g struction. Including the proparsuon ggy pg ggggg, g g9, gggg, ,,

• and implementation of procedures 34 m 8e31. Aer. 3. tsee: 34 m eT7o. Aor.

necesqto guide the effort.standard (g) Con /orMance'with-the (ILJ.3.1)l 33. tese:

tese7. 38 m totes.

Dee. 34.1970: se mJune asse.37.Pee.teto: 2s. 38 PR

.L J. ,. wm a) 2WmMLThsl tett: as PR 4sst. Mar. 13. 1971: as PR

" " -- -- .1p301.seet.11.teft! .

foggy _--*- ftensue9peeswee*uftr plant operaung  ? EerTossas. Norm: Per addtuonal Puenar.

Restarum estations affectans 6 50.34 see tne Eme>6MfgPshall include an evalua. tast of cPR soeuens Affsetse in une Ptnetas uon of the facility against the Stand, Aids section of Uus meunte.

ard Revtew Plan (SRP) tri effect on May 17.1982 of the SRP revtston in IM W ' I M''"

effect sta months prior to the docket to ro6emeen d h mese.

date of the application, whichever is "*8I" # ~^ ~ ~ P " ***

Later.

' (11) Appifcations for !!ght water (a) An application for a permit to cooled nuclear power plant construc-t!cn permits. mattufacturing !! censes.shal! construct a nuclear power reactor Irtefude a description of the pre-and preilminary or final desten accro- 11minary design of equipment to be in.

vais for standard plants docaeted after stalled to maintain control over rac!o-May 17.1982 snail include an evalua-tion of the facility against the SRP in act!?e materials in gaseous and Ilquid effect on May 17,1982 or the SRP re- effluents produced duttng normal re-vision in effect six months ortor to the actor operattons including espected l doctet date of the appilcation, which- operational occurrences. In the case of i an appilcation filed on or after Janu.

ever is later. '

(2) The evaluation required by this' ary 2.1971. the applicadon snail also identify the design OOjectives. and the section shall include an identification and desenstion of all differences in ' means to be employed, for 1teeping desien features, analytical techniques, levels of radioactive material in ei-and procedural measures proposed for fluents to unrestricted areas as low as a facility and those corresponding fes. Is reasonably achievable. The term "as tures, techniques, and measures given low as la reasonnely achievable as in the SRP seceptance criteria. Where used in this part means as low as is reasonably achievable taking into ac-such a difference estats, the evalus. count the state of technology, and the tion shall discuss how the alternat!*e proposed provides an acceptable economim of improvemente in relauen method of complying with those rules to benefits to the public health and or regulations of Commission. or por- safety and other societal and soetoeco-408 126 Revision A 06/01/84 i

l l

t

j 30.3d Title la "w th) A construction permit will consti. tions. The technteal specif! cations will tute an authortzation to the applicant be derived from the analysen *M ==*1 to proceed with construction but will untion included in the safety,ang not constitute Commission approval of -- anu amenqmsmaPe sub-the safety of any design feature or M pursuant to i 60.34..The Com.

spectitcauon unless the applicant spe. missioTaliayTacluile such addluonal cifically requests such approval and technical specificadons as the Com.

such approval is incorporated in the mission finds appropriate.

permit. The appilcant. at his oct!om (c) Technical specificauons wl!! In.

may request such approvals in the clusie items in the following cat $rtes*

construction permit or, from time to (1) .Te/ety limits, limiting se/eir time, by amendment of his construe. syste.m settings, and limiting control tion permit. The Commission may. in settings. (t)(A) Safety !!mits for nucle.

, its discretion. Incorporate in any con. at reactors are limits upon important, struction permit provtsions requiring process variables which are, found' to the applicant to furrush periodle re., be necessary to reasonably protect the ports of the progress and results of re. Integrity of certaan of the physical search and development pre, grams de. barrters which guard against the un.

signed to resolve safety quesuons. controlled release of radioecurity. If (c) Any construction permit will be any safety limit is exceeded the reac.

subject to the limitation that a license ter shall be shut down. The Ilcensee authortsing opersuon of the facility shall noufy the Commissiom review -

will not be issued by the Commission the matter and record the results of until (1) the app!! cant has submitted the review, including the cause of the to the Commisstom by amendment to condluon and the basis for correcuve the applicattom the complete final action taken to preclude reoccurrence.

safety analysts report. portions of Operation shall not be resumed untG which may be submitted and evaluat* suthertsed by the Commission.

ed from time to time, and (2) the Com.

(B) Safety limits for fuel reprocess. .~

mission has found that the final ing plants are those bounds wtuun desten provides reasonable assurance which the process vertables must be that the health and safety of the maintained for adequate control of

• pub!!c will not be endangered by oper* the opertuon and which must not be atton of the fact!!ty in accordance with escoeded in order to protect the intes.

the requirements of the !! cense and rtty of the physical system which is the regulations in this chapter. designed to guard stamst the uneen.

<5ec. 188,58 Stas. 3$$; 42 U S C. :::St trolled release of radiosctivity, if any (M m 12s13. Cec. 3s.1942, as amended at safety limit for a fuel reprocoesing 31 m Inso. Sept. 30. tese: 35 m 3318. Diant is exceeded, corrective action

• Mar. J1.1970: 35 m se44. Apr. 28.1970: 33 shall be taken as stated in the technt.

FR 11441. July 7.19701 cal specification or the affected part of the process, or the entire process if f 50.34 Technteel specificauens. required. shall be shut down. unless (a) Each applicant for a license such action would further reduce the "

authortsing operauen of a production margin of safety. The lleensee shall ,

or uullsatton facility shall include In notify the Coraraission. revtew the his applicadon proposed technical matter and record the results of the specificauens in accordance with the review. Including the cause of the con.

requirements of this sectiom A sum. dition and the basis for corrective mary statement of the bases or rea. action taken to preclude rerecurrence.

sons for such specifications, other If a portion of the process or the than those covering administrstive entire process has been shut down, op.

controls, shall also be included in the eradon shall not be resumed unul su.

app!!cauon. but shall not become part thorised by the Commission.

of the technical specifications. (!D(A) Limiting safety system set.

(b) Eacn license authortsing oper. Lines for nuclear reestors are settings atton of a production er ut!!!sation it. for autornatic protective devices reist.

c!!!ty of a type desertbed in I $0.21 or ed to those wartables having signifl.

I $0.:: will include technical specifica. ennt safety functions. Where a !!mit.

410 ,

l l

C6/01/84 127 Revision A

'em

{ 50.34 Chepter I-Hucieer Regulatory Commission any remedial action permitted by the Ing safety system setting is specified technical specifiradon untti the condt.

for a variable en which a safety !!mit has been placed, the setting shall be so ' tion can be met. In the case of either a protective nuclest reactor or a fuel reprocessing chosen that _*"h*Ha plant, the liernset shall notify the E action will correct the abnormal situs. Commission. review the matter. and tion before a safety umit is azeeeded.

II. during opersuon. tr.? sutomatic record the results of the review. In.

cluding the esure of the condition and aatety system does not funct!sn as re.

quired, the licensee shall take coro. the basis for corrective action taken to preclude reoccurrence.

I priate action. which may include sjlu, t.

down the reactor. He shall noufy (3) .TurorH24 ace reoutrementa. Sur.

tlin the Commission, revtew the matter veillance requirements are require. '

and record the results of the review. ments relating to test. calibrstfon. or '

including the esuse of the cond!Uon inspection to assure that the necessary I

and the basis for corrective , action . quality of systems and components is maintamed that factlity operation will taken to preclude reoccurrence.  !

(B) Limitint control settings for fuel be within the safety !!mits, and that reprocessing planta are setungs for the limiting conottions of operaMon '

automatic aJarm or protective dwices will be met.

• related to those var 1& Dies having sig. (4) Design features. Dealen features nificant safety functions. Where a to be included are those features of .

limiting control setting is specifled for the facility such as materials of con. t a variable on which a safety limit has struction and seemetric arrangements.  !'

been placed, the setting shall be so which, if altered or modified. would chosen that protective action. either have a sigmficant effect on safety and automaue or manual, will correct the are not covered in categories described abnormal rituadon before a safety in paragraphs (c) (1). (2), and (3) of flatit is excee1ed. II. during opersuon. this mon.

the automatic starm or protective de. (5) Admsntstrafice controls. Adtnin.

vices do not function as required, the letratin controls an the provisions m. -

licensee shall take appropriate action launt to orgamsadon and manage.

to maintain the vartaoles within the asent. procedures, murdkwging, limiting control-setting values and to review and audit, and reporting neces.

reptir promptly the automatic devices sary to assun operation of the fac111ty r

' or to shut down the affected part of in a safe manner.

the process and, if required, to shut (d)(1) This section shall not be down the entire process for repair of deemed to modify the technical spect.

automatic devices. The !!censee shall fications incluced in any license issued notify the Commission. review the prior to January 16. 1969. A license m matter, and record the results of the which tecnnical specti! cat!ons have review, including the esuse of the con, not been designated shall be deemed I

dition and the basts for corrective to include the entire safety analysts action tak2n to preclude reoccurrence, Mport as techmeal speci!!cauona.

(2) Limittna condiffons for oper. (2) An applicant for a license author.

siton. Limiting conditions for oper. , tsing operation of a production or utt. ,-

atton are the lowest functional C358e. , lisation facility to whom a construe.

bil, tty,,gr performance Wyets of eqgu Won permat has been Lasued prior to ment required fer safe opersuon off January 18.1969, may submit technt.

thlTac111ty. When a limiting conditfori cal spectitcauons in accordance with for operation of a nuclear reactor is this secuen. or in accordance with the not met, the licensee shall shut down requireraents of thts part in effect the reactor or follow any remedial prior to January 16.1968.

action permitted by the techntent spece l .

(3) At the initiative of the Commis.

ification unt!! the condicon enn be sfon or the lleensee. any license may l met. When a Ilmiting condition for op. be amended to include technical spect.

l erstion of any process step in the

! system of a fuel reprocessing plant is i ficauens whlen would be of the scope required if aand newcontent 11 not met. the tieensa shall that part of the otersuon or followshut down eense were bems issued.

i 411 i

Revision A 128 06/01/84

- - - - - - - , . , , - ,. ..-_ym_._ _. .-v,- - , . - - , - - - - , - - . -

o

• Chapter 1-Nucleet Regulefery Commission j $0.42 8 saas inellribility of certain applicanta. (d) Any applienb!c requirements of Any person who is a clusen, nacon. Part 51 have been asusfled.

al. or agent of a foreign country, or (21 m 388. Jan.19.196s. as amended at 3e any corporation. Or other enuty which m 12731. July 7.1971: 3e m :s:79. July the Commission knows or has reason 18.1974: 47 m 13754. Mar. 31,19821 to believe is owned, controlled, or .

dominated by an allen. a foreign cor. 8 54.41 Additieaal standards fee clean 504 parauon or a foreign government, lleoness.

shall be ine!!stble to apply for and In determining that a class 104 !!.

. obtairs a license. cense will be issued to an applicant, asec.181. as amended. Puo. 1.83-703, es the Commission will. In addition to ap.

Stat. ses (42 USC. 2:ott sec. 201. as plytng the standards set forth in amencet Puo. I. s3-438, as stab 1*t3 (42 150.40 be guided by the following con.

• U.S.C. 5841H ,, . siderations: .

. I:t m 355. Jan.18, itse, as amended at 43 f a) The Commission will permit the FR 8924. Feo.17.19781 widest amount of effective medical

, -~ theracy possible with the amount of I $8J9 l'shile insemet6en>f ab - - - - special nuclear matertal available for Applicatter.s arid documents submit. such purposes.

ted to the Commission in connection (b) The Commission will permit the with app!!cauons may be made avalla. conduct of widespread and diverse re.

ble for public inspection in accordarice search and development. .

with the provtssons of the regulauens (cJ An sopilcauon for a clans 104 op.

contained in Part 2 of this chapter. eraung license as to which a person **

who intervened or sought by timely SrMDAADs ros I.IcENsss AarD written nouce to the Commission to ConsTaccT10w Panserrs Intervene in the construcuon permit

$50.80 f*em m saandarda, proceeding for the facility to ootain a determinauon of anutrust consider.

In determining that a license will be allons or to advance a jurisdictional Lasued to an appilcant, the Commis* basis for such determination has re-ston will be guided by the fo!!owtng quested an anutrust review under sec.

considerations: tion 108 of the Act within OS days (a) The processes to be performed, after the date of publ! cation in the the operaung procedures, the fact!!ty PtnznAL Ras:sfun of notice of f!!!ng of and equipment the use of the fact!!ty, the application for an opersting !!.

and other technical specifications. or eense or Cecember 19.1970, whichever the proposals. In regard to any of the is later. !s also subject to the provt.

foregoing collectively provide reason

• acle assur.nce that the appilcant will stons of 150.4:tbl.

comply with the regulauons in this (42 U.S.C. 2132 2135. ::39 chapter. Int.luding the regulations in (*1 FM 235. Jan.19,1958. as amenced at *3 Pitt 20. and that the health and m 19sso. Dee. 29.13701 of the public will not be endan.

, I WI

• Additlenal standerde for etans 103

• tbl The applicant is technically and II""

f!nanctally qualifted to engage in the In determining whether a class 103 proposed activtues in accordance with Ilcense will be issued to an appilcant, the regulauons in this chapter. How. the Commtssion will. In addidon to ap.

ever no considerauon of financial piring the standards set forth in qualifications is necessary for an elec. I 50.40, be gulded by the following tite uulity app!! cant for a lleense for a consideradens:

production or uultsauon fact 11ty of the (a) The proposed activttles will serve type described in i 50.21(b) or i 50 :* a useful purpose proportionate to the tc) The issuance of a lleense to the quantitles of special nuclear matertal applicant will not. In the opinion of or source matertal to be utillsed.

the Commission. De inimical to the (b) Due account will be taken of the common defense and securtty or to the advice provided by the Attorney Gen.

health and safety of the public. eral, pursuant to subsection 10$c of 413 05/01/84 12e Revision A

• o ise Chaptw l-Weleet Repletery Commission ] 30.44 (2) A combustible gas control system the general requirements of Cntens is a system that operates after a LCCA

41. 4% and 43 of Append!x A to this to maintain the concentrauons of part. If a purge system !s used as part combustible saaes within the contain. E of the repressurtsation system. the ment. such as hydrogen. below flam.

purse system shall be designed to con, mability Umits. Combusuble gas con.

form with the general requartments of trol systems are of two types: (1) Sys.

Cnterta 41,42, and 43 of Appendix A tems that allow controued release -

to t!us part. The containment shaU from contamment. through filters if

• not be repressurtzed beyond 50 per. necessary. such as pursms systems I cent of the contamment design pres. and repressurization systems, and (11) '

sure. systems that do not result in a sigmfl. '

(g) For facilities with respect to cant Miense from contamment such as.

which the conce of hear:ng on the so, ,

p!! cation for a construction.per=st was . Mcombmers.(3) A purgmg system as a system for published on or before December 21 the controlled release of the contam.

1944,if the combmed radiation dose at ment atmosphere to the environment the low population zone outer bound. through futers if needed.

try from purging (and repassurma. (4) A repressuntation system is a tion if a repressurtsstion system is pro. system used to dilute the concentra.

vided) and the postulated LCCA calcu. uen of combustible gas wittun contam. L lated in accordance with 1100.11(aX:) ment D adding inwt gas or air to the  !

of this chapter is less than 23 rem to containment. Dilution of the combus. T the whole body and less than 300 rem table gas results in a delay in time' to the thyrold. only a purging system until a flammable concentration is ..

is necessary. provided that the purging Nached and pennits fission product system and any filtration system asso. decay. Operauon is limited to a con-etated with it are designed to conform '"A""'"I "8"8'"U8""*" '8 N D'f*

with the general requirements of Crt. unt M the matamment design one.

terts 41,42 and 43 of Appendix A to sure. A purstng system is normally this part. Otherwnse, the facility shall part W t e Monsunsauon system.

te provided with another type of com- (ase.141. as assended. Pua. L 83 703, es bustible gas control system (a repres. Stak M s #42 U A C. 12 cit sec. 201. as sunsation system is acceptacle) de. assended, ha. L 8348. Is Stat 1342. MD.

signed to conform with the general re. 1. 94 78. 89 Stal 413 (42 USC. $s41H Quirements of Cntens 41,41 and 43 of tes FR 50183. Cet. 27.1918. se amended at '

Appendix A to this part. If a purse ** I I 30*00* OI*'80II system is used as part of the repressur.

zauon system, it sr.all be dos:gned to 8 54.43 Standarea for construeuen pet.

conform with the tenersi recurre. ma.

ments of Cnteria 41. 41 and 43 of Ap.

An applicant for a Ucense or an pendia ment shallA to tlus notpart. The contam. amendment of a Ucense wno proposes be repressun:ed beyond 50 percent of the contamment to construct or ut!Uzauon 14ter a facility producuon viu be initially or design pressure.

(h) As used in this section: (D Deg. granted a construcuan permat. If the ,

radauon. but not total failure, of appucadon as in conformity with and emergency core coollag funcuoningll acceptaale under the entens of 50.31 through 30.33 and the stand.

means that the performance of the ards of Il S0.40 through 30.43.

emergency core cooling system La poa.

tulated, for purposes of design of the 9 34.44 Asewsamee emens fee emergwy l

combusuble gas control system. not to me e n.g . fw us= emer ,

meet the uceptance enteria in i so.4e neslear power reesters.

l and that there could be localised clad (a)(1) 22 cept as preended in paru.

i Inclung and metal. water rescuan to the extent postulated in paragraph Idi graphbelling Iax2) and and (3) of this section.

pressurtzed light.

of this rection. The degree of perform. each water nuclear power reactor fueled ante degradation is not postulated to with uranium oxide penets *1thm cy.

be suf ficient to cause core meltdown.

417 l

I Revision A 130 l 06/01/ N

i 50 4 Title 10 g ,,,

11ndrical Z1rcaloy MaMirW shall be provided with an emergency core cool.

complete IL The Director of Regggg.

Mon of the Atomic Energy Commtasson tag system (ECCS) which shan be de.

alsned such that its calculated cooung shall have caused notice of such a re.

performance following "'"larad loss.

quest to be puhushed promettr m the PEsm&L RastsTsK such houce snagg of. coolant acetdants conforms to the have prwetoed for the subalaston o(

criterta set forth in paragratta (b) of comments by interested persons this section. ECC3 coottnr perform. wrthin a time pertod establianeo my ance shall be calculated in accordance the Director of Regulacon. If. upon with an acceptable evaluauon modet, rettewmg the foretoms submtsatort and shall be cale . lated for a number the Director of Regulation conc!uce.s of postulated loss.of coolant acc:denta that good cause had been srtoms for of different sizes. locscons. and other an extension. he may have extenoco

• 1 properties sufficient to provide assur. the six month period for the shortest ance that the entire spectrum of pas. additional time which in his judgment tulated loss-of coolant accidents is cow. will be necessary to enable the licensee ered. Appendia E. ECCS Evaluation to furnish the submissions required by 3dodels sets forth certain required and paragraph (ax2X11) of this section. Re.

acceptable features of evaluauen quests for extenatons of the 312 montri models. Conformance with the criteria perted submitted under this subpart.

set forth in paragraph (b) of this sec. graph will have been ruled upon by tion with ECCS cooling performance the Director of Regulation prior to ex.

calculated in accordance with an ac. piration of that pertod. e ceptable evaluation model, may re. (iv) Upon subnussion of the evalua.

Quire that restricticas be intposed on tion required by parsgrach (aM2X11) of reactor operation. this section (or under parsgraph (2) With respect to reactors for (&M 2X ul). Lf the six month period as wluch opersttag licenses have preft. extended) the faculty shall continue .

ously been tasued and for which oper. or commence operation only withm ating Ilcenses may issue on or before the limits of both the proposed techna.

December 28.19741 cal spectfications or ucense amend.

(1) The time within which actions re. menta subautted in accordance with quired of pertnttled under this part. this paragraph (aX2) and all technical graph (aH2) must occur than begin to specifications or license cond!: tons run on February 4.1374. prevtously imposed by the Atom:c (ill Within six =ontna foUowing the herty Commissaan. Including the re.

date spec:fied J1 parsgraph tax 2X1) of autrements of the Interim Po11er this sect; ort an evaluation La amord. Statement (June 29. 1971. 36 m

• ance with paragrsch (aX1) of this Sec. 12:48) as amended December 13.1971.

taan shan have been submitted to the 36 FR 24082).

Ctrector of Regulacon of the Atonne (?) Further restrictions en resetor Energy corr.massion. The evaluation oceration win be unposed Lt it is found shall have been accompanied by such that the evaluations submitted unoer -

proposed changes in technical specifl. paragraphs (aX2) (11) and (111) of this cations or 11 cense amendments as may section are not consistent with part.

be necessary to Drtas reactor oper. graph (SX1) of this section arid as a atton in conformity with paragraph result such restrtettons are required to (ax1)of this section. protect the public health and safety.

(111) Any 11censee may have request. (vt) Exemptions from the operating ed an extenston of the six. month requirementa of paragraph (ax2Xivt perted referred to in paragraph of this section may be granted for (aH2X11) of this section for good cause. good cause. Requests for such exemp.

Any such request shall have been aut> tion saan be submitted not less than matted not eas than 48 days prior .. 48 days prior to the date upon which expiration of the six month period, the plant would otherwise be required and shall have been accompanied by to operste in accordance with the pro-affidavita showtng precisely why the cedures of said paragraph (aM2Xtys of evaluation is not complete and the this sectiert. Any such request shall be m:nimum time believed necessary to f!!ed with the Secretary of the Com.

418 i C6/01/84 131 Revis4cn A

- - -_ _..___i_.,z_.___ ._

o . \

l l

l m

j $0.44 Chaptee I--% leer 2eguietory Commission I

mission who shan cause notice of its ed to occur, the inside surfaces of the l recelot to be puolished promptly in cladding than be included m the oxi- r the FzosmAL RaczsTza, suca nouce dation, besmnms at the emiculated snali provide for the suh=ta=* of time of ruptum. Cladding tNearn=== g comments by laterested persons before oxidation means the radial dis-withm 14 days following PassaaL Ras- tance from inside to outalde the clad-tsTan pub!!caden. The Director of Nu. ding. after any eniculated rupture or -

i clear Reactor Regulation shall submit swelling has occurred but before sis- .-

his flees as to any roguested eXemp- nif1cantoutdattmg, Whggg the cElCulat. f uen withm f!ve days followmg expira- ed condluons of transient pressure and con of the comment period. temperature lead to a I,rediction of (vt!) Any request. for an exemption cladding swelling. with or without sutmutted under paragraph tam 2Myt) cladding rupture the unexidised clad.

I of th:s section must snow, with appro . ding thicameas shau be defined as the t

priate affidavits and techancal sutmus. cladding cross secuonal area, taaen at saons, that it would be in the public in- a horizontal plane at the elevation of

}- '

l terest to allow the licensee a specified the rupture. If it occurs. or at the ele- l -

additional period of time within which vtuon of the highest eladding tem- 1 to alter the operauen of the facility in perature if no rupture is calculated to l the manner required by persgraph occur. divided by the average circum-tam Mtvl of this section. The request forence at that elevation. For ruptured shan also include a discussion of the cladding the circumference does not alternauves avadable for estabilshing include the rupture opening.

(31 Mestemm Apdrogen generettos compliance with the rule. The calculated total amount of hydro-(3) Construction permits may have been :ssued after December 28, 1973 gen generated from the chenucal reac-but before December 28. 1974 subject uon of the cladding with water or to any applicable conditions or restric. steam than not exceed 0.01 times the uans imposed pursuant to other regu- hypothetical amount that would be i

.lacons in this chapter and the Interim generated if all of the metal in the Acceptance criteria for Emergency cladding cy!!nders surrounding the Core Coollag Systems pubushed on fuel, escluding the cladding surround-June 29,1971 (36 FR 12248) as amend. Ins the plenum volume. were to react.

(4) Coodsede geometry. Calculated es (December 18.1971. 34 FR 24082):

Prottded, hoteeter. that no operating changes ta core geometry snall be j

license shau be tasued for facultles such that the core remame amenable constructed in accordance with con. to cooling.(S) I.ang term cooling. After any es!-

i struction permits tasued pursuant to this paragraph unless the Commtulon culated successful mitia! operauen of l deternuses, among other thmss that the ICC3. the calculated core tam-

! the proposed factuty meets the re- perature shan be maintamed at an ac-t curementa of paragraph (aX1) of this ceptably low value and decay heat l

see .on. shall be removed for the extended

[

i F H1) Peak cladding temperature. l pertod of time required by the long-11ved T3+ calculated maximum fuel element) core. radioactivity remairung in the cAdmg temperature shan not esceed , , .

2:0J' F. (c) As used in this section: (1) Isas-C) Me21 mum cladding ersdation of coolant acendents (1.0CKs1 are hy-pothetical accidents that would result The calculated total oxidattoft of the tisdding shall nowhere exceed 0.17 front the loss of reactor coolant at a tirses the total eladding thachness rate in excess of the capaDality of the reactor coolant makeup system, from beiore oxidsuon. As used La this suD. breabs in pipes in the reactor coolant paragraph total oxidation means the total thielsness of cladding metal that pressure boundary up to and including aculd be loestly converted to otide Li a break equivalent in asse to the all the oxygen absorbed by and react. double-ended rupture of the largest ed with the cladding locally were con. pipe in the resetor coolant system.

l verted to stoichiomettle stressuum (21 An evaluauon model is the eticu.

dientee. :t cladding rupture is eticuant. tauonal framewers for evaluaung the 419 l .

i 132 Revision A l 06/01/84 i

i

e o t .

I 34.47 TlHe 1? ,

behartor of the reactor system during  !!nding wul consutute a rebuttable a postulated less of coolant ***at pronusaption on questions of adequacy (LOCA). It tacaudes one or more com- and implementaden capabWty. Emer-puter programs and all other informa. gency preparedasse excretses troquired uen ama== mary for applicauen of the by paragraph (bx14) of this section calculadonal framework to a specifle and Appendia E. Secuen P of this LOCA. such as mathemaucal models part) are part of the operadonal in- .

used, assumptions incluoed in the pro- spection process and are not required grams, procedurs for tretung the pro- for any inattal licensing decision, grant input and output informauon. (b) *!1 e onette and, except as pro-specificauon of those port!cas of anal

• vtded in parasTaph (d) of t.ts SecuCA.

ysm not included in computer pro- offsate emergency response plans for

,- grams, values of parameters, and all nuclear power reactors must meet the otaer informauon necessary to specify followtas standards:'

the calculanonal procedum. (1) Primary responstbulties for emer-(d) The requirements of this section sency response by the nuclear fact!!ty are Lrt addluon to arty other require.

ments appucable to ECC3 set forth in Licensee and by State and local organs.

sations within the Essergency Flan-this part. The criterta set fonh in ning Zones have been assigned, the paragraph (b). with cooling performa emergency responsibuities of the Var.

ance calculated 2 accordance M an tous supporting organisations have acceptable evaluauert model art in implementation of the general re- been speetlicagy amenhtiahed, and each principal response orgarusation has autrements with respect to ECC5 cool staff to respond and to augment its ing performance design set forth in taltial response on a conunuous basis.

tha part, including in parch Crite*

rien 33 cf Appendia A. (2) On sht!t faculty licensee respon.

anbulues for emergency response are (30 PR 100s. Jan. 4.1Mt as asneesed at 30 unambiguously defined, adequate yR mal. Jutr 28 tMt 40 yn Mse. 3sar. 3. staffing to provide laattal facility aces.

IM33 dent response in key fumettonal areas

!s maatamed at all times. timely aus-

"*'"W***' mentauen of response encabalues is (aX1) Escept as provided in para

• avadable and the laterfaces among graph (d) of this secuen, no operaung various onstte roeponse actiftues and license for a nuclear power reactor will offatte support and response activtues be tasued unless a ficcing is maos by are specuted.

NRC that there is reasontale nasur. (3) Arrangements for requesting and anee that scequate protective mens

• effectively using assistance resour.'es ures can and wW be taaen in the event have been made. arrangements to ac*

of a radiolor. cal emersency, commodate State and local staff at *he (2) The .VRC will Dase its finding on Licensee's near site Emergency Cper-a review of the receral Emersency attons FacWty have been made, and Management Agency (FEMA) f1ncings other organissuona capable of aus. ~

and determinations as to whether menting the planned response have State and local emergency plans are been ident1 fled.

adequate and whether there is reason.

able assurance that they can be tsaale. (4) A standard emergency classif!ca-Hon and action level scuame, the bases 11ented. and on the NRC maaanne ent of which include facWty system and as to whether the applicant's onsite ei!!uent parameters is in use by the ernersency plans are adequate and whether there ta reasonable assurance nuclear faculty !!censee and State and that they can be implemented. A local response plans call for relisace FEMA finding WW primartly be based cn a review si trle plans. Any of tr in. 3, ,,,,,,,, me are by sped.

, 3,gggg,,,,,; pg3gg,ggy g formauon already avadable to FEMA enutted 'Cnteria for Preparauen and Eva&-

may be considered trt asseestag wheth* uscen of naaselegiens Emereener Reseense er there ts reasonable assurance that Pans and Pweareeness a supeert of Nucie-the plans een be tmplemented. In any at Power Planta-ter tasertan Use and Cam-NRC licensing proceeding, a FEMA ment *,Januarr tree.

4::0 . ,

1 06/01/84 133 Revishn A i

. .a -

• e em j 3057 Chapter L-Nucleer 2eguietery Commissiert authertsed 97 the Coranuasion uoon request punuant to l So.JSat ax:xu).

Poemedes te 130.88a ' For purposes of this ree*alation the gro.

' (Reserved] pened IEEE 279 teense **ta effset" on e Ceasementa Walsh are emesseted te tae Aueust 3o.1968. and tae revised (mue IEEE E remeter emelant system and are part of tas 2TMST1 tesame '*1a effset" on June 3.1971.

reasser esonest ersomre soundary esf*neg Cesses mar to esta&aed from me faststute in ise.3tes need met ases tasse reeutre. of Elastriani and Etestresses Enesseers, mensa, prootese: Casted Emanneertae Center. 348 East 47ta I' ial In tae enst of W fauun of Street. New Yorts. NT toe 17. A 4097 la sead.

I tae component eunes normal Master om. tale fer tasescues as tae Camauessen's tae meetw ma W anut eews and Pueue Doc gg,ggggg,,ument Aseen.1717 E Street NW*

,gg ese6ee dews la na oreerir meaner ' ensumane ,

• ' Where am neouemusa for a osastrueuca ,

ontr. or peruut le eineutted ta four parts nursuant (t) The amusensat u or saa de usisted* to tne arensons of I alma as and Junoars I from tae reecser oselaat ensam er two P af Part 2 of taus caseter. "tae fornial >

ename steta cisme. mta osen. w ome mu scieses decaet ease of the neouemuos for a coa.

serueues swaa" for ourse es er tais me.  !

i ama cae me essasse of mutamasse assumuss einer esena. Emaa oesa a'**

'*'"end.uon sankt be tae ease of cocaeums of sne m.

e tae omw new is open, na caesum formances tame must to suma taas. ta tae event of see, w g g, r,equ,a, red by i 2.101ca.13 (: or (2)*

talanes fauure of tae e durtas 8 $8.88 Caseeressa of constrasmen perma asemaa remeter eserausa. ease emin to. to licennet er ammendment e( tisemen,

, masas opernale ame the roaster eas as anut ereerty Upon onepletica of the construction ,

deve,,,and,,,esseed n,, ,

, ,,,, , ,,,ee,ws, , g , la, ,na,,ia.gor3,alteracon of a fact 11ty, in compil.

tae reester osanaat measus erasen emir. anee with the terms and conditions of s ceases may to estaaned from tae Assert. the COSetructioS perlait and SuDjec*. to ean Somety of 3deemasiani h any neceseary testing of the facility Omated himaa=tne Center. 348 East 47ta for health or safety purpenes the at. New Yora. NT 10e17. Caeses are-a araus. raammismann will, ga the annonce of ,

tie for tasessmen as tae <*

Putus Desument messa. 1717 3 3L N W = good cause shown to the contrary '

• • • "'*8'*"' EE" !asus a lleense of the class for whics T RAS and 6 Caes addends W the construcuan pertist was lasued or an appropetate ament.mnent of the !!.

3d" '* *,"'.ae n'*

• e, usuanse s.d nasn d . m t=e - =ar x '

me.u.s an , m.,, 4a (See. ISS. 64 Stat. $$$; 42 !ES.C. ::38) aner taer an Laemenase Dr niennae a parserson (t) of tat.s soeuca. Amesada to (!! PR 388. Jaa.19. Its1. as senensed at 34 tne AAME Coos asuee after tas Jumater II 1I001*EukF IT IIT01 1777 amannan are ceaAleered to De *ta f 34 37 !asennes of opeasting license.'

offset" or"effocuve" af*ar tae ease of SuBU.

y ention of the adoefula and after tBer are 13 (a) Mrsuant to 1 W46 an ocerst:.ng corsernsed or reference a pernersea st; ef license :nay to tasued by the Cat:=s.

uus soeuse. saan up to the full terin author. nd By

'Por A8ME Case Eatcome and Addenes teued prior to tae W1 ster 1911 aseensa. I 30.31. Upon finding that; W Construccon of the fac:llty '.*.as tae Case Edlues and Adaenes secuenste te ,

tae ceaseeems to severned my tae steer or been sutsta tually completed, in con.

eentreet date for tae sesseeeest, set the formity wit.1 the construction permat austrast ease for tae ausaear escret erssen, and the applicauen as amended. the Per tas 1rtster tyn aseeses and esans, provtstens of the Act. and the rules euset seitsens and adesses tae mothed for and regulations of the Coensuasaca:

desertananas the aseuenste Code celuees ame aseeses la omstaased ta Persernes NCA and 1160 of asemen IIIof the AsME Case. - 'The t"=-a= mar tasue a prestatenal engME Caes esses valen aan toen de. eseratag usemas pursuant te tae rp'da.

teraused anstante ter use 37 tae Commas. taens .4 taas part la effset se Maru. 30.

ase staff are usted un SC Reeulasary lato, for aar fae11tr for wanea a nettee of Quade 1.84. "Caes came Assestastutt asenne se as noeuencea for a PrunasenaJ AstEE Serusa !!! Desen and Peartsmusa* operatame usease or a nouse of presemed 3 Case suaaa *8

• prenaisons op.rauns uanse em and NRC .Re.eWaserr car AsME soeuQuade e :: 188. ==a"r Ms. es er sofon taas dase.

suomae*

Can terma. Ames Tn use of etaer Cue eues ear se 437 g Revision A

! 06/01/84 f

.J

- - - , - , , ., , - , - _ , . , - - - .m -n__ _ _ _ , , , , , _ , _ , . , ,, , , , _ _ , . . _ _ _ , , ,, _ _ , _ , _ _ _ , , _ , _ _ . _ _ , _ , _ , _ _ _ _ , _ _ _ _ , , , , , _ , , _ _ _ , _ _ , _ _ _ _ _ , , . _ ~ _ , _

m . _ _m __- . . _ _ _-m - . _ _ _ . _~

g o

• l e t 1

} 30.38 mie 10-En.,9y (2) The facility will operate in con. this section as to wnich there is a con.

formsty with the application as troversy in the form of an init!al deci.

amended, the provtssons of the Act, slon with respect t.3 the contested ac.

and the rules and regulauens of the ttvity souant to be authertsed. The Dt.

Commission: and rector of Nuclear Reactor Regulation (3) There is reasonsbic assurtnee (1) will make findings on an other matters that the activities authorised by the specified in parss sch (a) of this sec.

Opersting license can be conducted tion. If no party opposes the mouon.

without endangering the health and the presiding officer will tasue an

, atlety of the puntic, and (11) that such order pursuant to l 2.730(e) of this activttles will be conducted in comp 11 chapter. authortzing the Directer of ance with the regulations in this chap- Nuclear Reactor Regulauon to make

~

ter. and approprtste findings on the matters *

(O The applicant is t.echnically and specified in parastaan (a) of tnis sec. .

financtaur quallfled to activttles authortsed by,enesse in thequested the operating tion andopers'.lon.

to imue.a lleense for the re. '

license in accordance with the regula.

Mons in this chapter. However, no (3s m sats. :s.u. 31. tote, as asiensed et striding of financial cual!!! cations is 3s m M. Apr. 28. 2M 31 m 1ts13. June II necessary for an electric utility appil. hI 38h 9 g 3 1 2.

cant for an opersung Ucense for a pro. goggj duction or udlisation facility of the type desertbed in 130.21(b) or I $0.22. 9 50.38 Heartees med revert et une Ademo.

(5) The applicable provtstons of Part ry Cassannes se Reneeer 5efesserde.

40 of thas enacter have been satisfled; gy ,, ,

(8) The issuance of the license will tion Domit or an oms Usense for not be inimical to the comunon defense a facility which La of a type described and security or to the health and in l S0.21(b) or i 30.22. or for a testing -

safety of the pub!!c. facility, shall be referrht to the Adyt.

(b) Each operating license wtU in. sory Comenittee on Reactor Safe.

clude approprfate provtalons with re. guards for a review and report. An ap.

spect to any uncompleted items of pil>:suon for an amendment to such a construction and such limitations orl constmetton permit or opersurig H.

conditions as are required to assur,l oense may be referred to the Advtsory that operstion during the perted of f Committee on Reactor !!ahst.ards for the completion of such !tems wtll notj review and report. Any recort Jhall be endanser puolle health and safety, i made part of the rword or the applies.

(c) An app!! cant may. In a esse uen and avadsele to the puDilc. except where a hear:ng ts held in connection to the extent that security claastfles.

with a pending proceeding under this Won prevents disclosure, section make a motion in wrtung pur. (b) The Commission will hold a suant to this paragrsch (c). for an op. hearing after at least 30 days notice -

ersung license authortsing low. power and publicauen once in the FussRAl.

testing (operation at not more than 1 Rsesseum on each applicauon for a percent of full power for the purpene construction permit for a production of testing the faci!!ty). and further op. or uullsation fac1Hty which Le of a erstions short of full power operation, type described in 130.21(b) or 150.22 Action on such a motion by the presid. or which is a testing feellity. When a Ing officer shall be taken with due construction permit has been issued regard to the rights of the parties to for such a facility following the hold.

the proceedtnss. Including the right of ing of a pubut hearing and an appilos.

any party to be heard to the extent Lion is made for an operstans Usense that his contenuons are relevant to or for an amendment to a construction the activity to be authertsed. Pr or to permit or operating !!eense, the Com.

talutig any action on such a motion mission may hold a hearing after at 1 which any party opposes, the presid. least 30 days notice and pubucation Ing off!cer shall make findings on the once In the FusanAt. Rsessena or, in matters spectfled in patssraph (a) of the abseries of a request therefor by 438 06/01/84 135 Revision A

--u.. a= _ _ _ , , , , _ , __ _ ,

l 4

p Chapter l-Nucteer Regulevery Commission l 30.70 any person those interest may be af- ments carried out pursuant tn para.

fected. may lasue an opertuns !! cense graph ta) of this section. These rec.

or an amendment to a construcuon ords shall include a written safety pertnet or operaung license without a evalusuon which provides the bases heartas, t.pon 30 days nouce and pub- for the determination that the change, E listion once in the Funesas. Ras stun test or esperiment does not involve an of las intent to do so. If the Commis- unreviewed safety question. The 11-seen finds that ne significant hasards consee shall furntsh to the appropriate .

conenders:lon is presented by an appil* NRC Ressonal Office shown in Appen.

cauon fcr an amendment to a con

• dix D of Part 20 of this chapter with a structio:t permit or operauns license, copy to the Director of Inspection and it may d*spense with such notice and Enfortement. U.S. Nuclear Regulatory publicauon and raar issue the amend. Commission. Washington. D.C. 20855. *

• ment- annually or at such shorter intervals (27 m 12tte. Dee. s.1ses, as ameneee as 33 as may be spectfled in the license, n m aste. June 12.1968: 35 FR 11441. Just report contamins a brief desertation 11.1Mo; 39 FM 10858. Mar. 21. tM41 of such changes tests, and espert.

ments. Including a summary of the l 3 64R h immes eM emprimenes. [ safety evaluauon of each. Any report t s)(1) The holder of a licenset submitted by a licensee pursuant to }

this permeraph will be made a part of , <

authertains operation of a production l t

or utilissuon facility may (1) make the public record of the licensing pro-coeques. In addition to a stened ortet.

)

chansee in the facality as desertbed in the safety analysis report. (11) make nal. 29 esp 6es of each report of chantes in the procedures as desertbed changes in a facility of the type de.

4 In the safety analysie report. and till) sertbed la 150.21ths or 150.22 or a tentlas facility, and 12 copies of each conduct sersbed tests in theor esperimenta safety analystsnot de'.

report report of changes in any other fact!!ty. l

, I i

without prior Commission approval. shall be filed. The records of chanses unless the proposed change, test or es* ' In the fasulty shan be maintained periment involves a change in thel untu the date of termineMon of the II.

technleal spostficaMoes lacorporatedl conse, and records of changes in proce.

In the lleense or en unreviewed safety' dures and records of tests and espert.

question. ments shan be raatntained for a period i (2) A c _ M change, test. or ex* of five years.

periment shall be deemed to Involvet (c) The holder of a !!c.ense author.

an unrettewed safety question (1) Ifl ! sing operation of a production or utt. {I the probabt!!ty of occurrence or thet !!sation facility who destres (1) a consequerices of an accident or mal 4 change in technical specif! cations or l

' function of equipment important tel (2) to mahe a change in the fact!!ty or safety previously evaluated In thej the procedures described in the safety

(

safety analysis report may be Ina l analysis report or to conduct testa or creased; or till if a poestbility for ans espedmonta not desertbed in the scendent or malfunction of a dillerenti s&fety analysts report, which involve '

type than any evaluated preytcusly in, an unrevtewed safety question or a the safety analysts report may be cre* chance in techmeal specifications, '

sted: or till) If the snarsin of safety sa shall submit an application for amend.

defined in the basis for any technical ment of his license pursuant to i 50.90.

spectflendon is reduced.

(bl The lleensee shall maintain ret.

t3e FR leste. Mar. St. tM4. as amenese at ords of changes la the facility and of 41 FR isees. Aer,19. IMG. 41 yn te30s, war 3. IMe:43 FR asias. Apr.18.18771 chanses la procedures made pursuant to this section, to the essent that such , Worsenus. Racones. Mom.

changes constitute chances in the fa. Nmncanne f cility as desertbed in the saf ety analy.

i ses report or constitute chanses in pro- , g,7, g,,,,,,,,,,,

cedures as desertbed in the safety analysts report. The licensee shall also ,

fa) Each licensee and each holder of mamtain records of tests and esperi- a construction permit shall permit in.

439 l

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, t l3C.90 TlHe 10-Ene'TY with the regulations in this chapter RavoCAT!cN. SesrER$ ION. McD!rtCA.

and will not be inimical to the Trow. AurwourNT or LICENsts ANO common defense and security or to the CoRsTRUCTION PsaMITs. EmsmosNc7 health and safety of the public. Orsaartons av Tus Consasassion (b) If the application demonstrates that the dismantling of the facility 8883# M pension, modif co.

and disposal of the component parts tien of Ilemas and cenetraction per.

will be performed in accordance with "'f*****-

the regulations in this chapter and A license or construction permit may will not be immical to t!te common de. be revoked. suspended or modified. In fense and security or to the health whole or in part for any maternal false and safety of the public, and after statement in the application for 11 notice to laterested persons. the Com. cense or in the supplemental or other t

massaon may, lasue an order author. statement of fact required of the 10 -

taing such dismantling and disposal, plicant: or because of conditions re-and providing for the termination of vealed by the application for license or the license upon completion of such statement of fact or any report.

procedures in accordance with any record, insoection. or other means, conditions spectiled in the order. which would warrant the Commission I:s m 9546. Oct. Io.1961, as amended at 32 to refuse to grant a license on an origt.

m m m 21.Iset) nal applicauon (other than those re-laung to i1 50.51. 50.42(a). and AnacrosstNT or LICENSE OR CONsTRUC. 50.43(b) of this part): or for fa11ure to T!6N Pramtr AT Rzeous7 or Hor.orn construct or operate a facility in ac.

cordance with the terms of the con.

3 $4.M Application far amendment of I;. struction permit or license, provided eense or eenstruction permit. that failure to make timely completion of the proposed construction or alter-Whenever a holder of a license or ation of a facility under a construction construction permit

• destres to amend the license or permit, application for permit shall be governed by the provt.

stons of I 50.55(bn or for violauen of, an amendment shall be filed with the or failure to observe. any of the terms Commission. fully desenbtng, the .

and provisions of the act, regulattons, enanges destree. and fonlowtng asTar 11 cerise, permit. or order of the Com-as appiscaole the form present:ed tor mission.

criemal scoifcations.~ i f 50.101 Retaking possession of special I $4J1 Isauence of amendment- nuclear matmal.

In determining whether an amend

• Upon revocation of a !! cense, the ment to a license or censtruction Commission may immeciately cause permit will be issued to the applicant the retaking of possession of all spe-the Commission will be guided by the cial nuclear matertal held by the 11-considerations whicts sovern the issu. censee. ~

,ance of Initial licenses or construction permita to the extent applicable and (21 m 355.Jan.19.1954, as amended at 40 sporcpriate. If the application in. m atse. Mar. 3.19731 volves the matertal alteration of a 11* 8 50.102 Commission order for operation cerned facility, a construction permit after reveession.

will te Lssued prtor to the issuance of the amendment to the license. If the W never tM Comnussion finds amendment involves a significant has. that the public converuence and neces.

sity or the Department finds that the ards consideration. the Commission production program of the Depart.

wt!! give notice of its proposed action pursuant to 12.105 of th chapter ment requires continued operation of

, before acting thereen. The notice will a production or utilissuon facility, the be issued as soon as practicable after I for which has bun revoked.

the application has been docketed. the Cornmission may, after consulta-Clon with the approprtate federal or

[39 m 13228. Apr.12.1974I state regulatory agency having furts-444 l

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Chenter I-Nveleer Reguletery Commission Port 50. App. A diction, order that pansession be taken compilance with the rules, rerutauons.

of such facility and that it be operated or orders of the Commission.

for a period of time as, in the juds. (c) The Commission may at any t!me ment of the Conunission, the public require a holder of a construction convenience and necessity or the pro- permit or a license to submit such in- j duction program of the Department formation concerning the add!Uon or stay require. or unul & !! cense for 09- proposed addition. the elimination or erstion of the fastlity shau become ei- propcoed e!!minadon. or the modifica-factive. Just compensauon shall be tlan or proposed modificauon of struc-paid for the use of the fact!!ty. tures, systems or components of a fa-

• etlity as it deems appropriate.

(40 FR 87s0. Mar. 3.1973)

P la war m m sats. Mar. 31. Im 8 54.193 Suspensson and

.or andeaal esnersoner ,

gryoacusamer . .

(a) Whenever Consress declares that a state of war or nauonal emergency 6 50.11e Vieleueam.

esists, the Comnussion. If it finds it An injunction or other court order necessary to the common defense and may be obtained prohibiting any viols-security, may, uon of any provtston of the Atomic (1) Suspend any license it has lasued. Energy Act of 1954, as amended. or (2) Cause the recapture of special T1ue II of the Energy Reorgsmaation nuclear material. Act of 1974. or any reguladon or order

. (3) Order the operation, of any 11- lasued thereunder. A court order may

.eensed facility. be obtained for the payment of a etyt!

(4) Order entry into any plant or fa- penalty imposed pursuant to section cility in order to recapture special nu- 234 of the Act for violation of section clear matertal or to operate the facill' 53,57,62.63.St.82.101,103,104.107 ty, or 100 of the Act. or section 20E of the (b) Just compensmuon shall be paid Energy Reorganimauen Act of 1974. or for any damages caused by recapture any rule. regulation. or order issued 2 , of special nuclear material or by oper- thereunder, or any term. condtuon. or ation of any facility. pursuant to this limitauen of any license issued there-secuen. under, or for any violation for which a (sec.104, es Stat. 939. as essended: 42 !! cense may be revolted under section UAC. 31383 ISS of the Act. Any person who wtll-(21 FR 388. Jan.19.1956, as amended at 39 fully violates any provision of the Act PR 11418. July 17.1970: 40 FR 3790. Mar. 3. or any regulation or order issued 19751 thereunder may be gut!ty of a ertme and, upon convtetton, may be punished

l. 3""* by fine or imortsenment or both. as i

1 54.109 Backlituns, provided by law.

Ial The Commission may. In accord. [40 PR 87s0. Mar. 3.1973. as amenced at 42 ance with the procedures specif!ed in FR 37:1.May 18.18771 this chapter, require the backf!tting of Arrerotess a facility if it finds that such action #

will provide substanual. additional protection which to required for the public health and safety or the Artsworz A--GewsmAt.Daston common defense and securtty. As used CaHealA rom Norma Powsm PLAwes in this section. "backfitung" of a pro- N # C8""

duction or uullsation fact!!ty means the addition e!!minadon or modifica- wraoeverson tion of structures. systems or compo-nents of the facility after the con- 8'FW8T8

struction permit has been issued. Nuefear Power Unit.

(b) Nothing in this section shall be 1,ses o( coolant Aesidents.

deemed to relieve a holder of a con- Senese Fatture.

struction permit or a license from Andelpated Coerational Cum.a.

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