Text

Discusses Incorporation of Items Into Westinghouse STS Deriving from NRC Confirmation of Generic & multi-plant Action Concerns from R Licciardo Review of Plant Tech Specs
	ML20236E297
Person / Time
Site:	McGuire, Mcguire
Issue date:	03/15/1989
From:	Licciardo R; Office of Nuclear Reactor Regulation
To:	Butcher E; Office of Nuclear Reactor Regulation
Shared Package
ML20236E296	List: ML20236E293; ML20236E297;
References
	NUDOCS 8903240078
	Download: ML20236E297 (4)
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k. . . . . # MAR 151989 MEMORANDUM FOR: Edward Butcher, Chief Technical Specifications Branch Division of Operational Events Assessment 1

FROM: RgbertB.A.Licciardo,ReactorEngineer(Nuclear) i Plant Systems Branch Division of Engineering and Systems Technology

SUBJECT:

INCORPORATION OF ITEMS INTO W STS DERIVING FROM NRC l CONFIRMATION OF GENERIC AND FULTI-PLANT ACTION CONCERNS 1 FR0ii R. LICCIARD0'S DP0 REVIEW 0F THE MCGUIRE TECHNICAL i' SPECIFICATIONS During a meeting with James H. Sniezek, the Deputy Director for NRR asked ! the writer to forward to your branch for implementation, those items of the l writers DP0 review of the McGuire Technical Specifications already confirmed I for incorporation into the Westinghouse Standard Technical Specifications (W STS). The subject information is contained within the attachment entitled

         ' Identification of Generic Items Confirmed for Westinghouse Standard Technical                           !

Specifications." Of the original 380 concerns, 220 items were selected by I Division of Systems Integration / Reactor Systems Branch (DSI/RSB) for review by j Division of Licensing (DL) and in the attached document 90 are finally categorized i as Generic, and 12 as Multiplant Action. Residual plant specific items numbering 45 are not identified in this document at this time. Thewriter'sreviewoftheMcGuireTechnicalSpecifications(seeReference1 in the attached List of References), reported 380 items of concern. Reactor Systems Branch (RSB) subsequently selected 220 of these for verification and these were sent to the Division of Licensing (DL) for that purpose (Ref. 2). DL responded to this request by memo dated May 28,1985(Ref.3). In its review, (Ref. 3) DL categorized three groups of conclusions: " Generic" (G),"PlantSpecific"(PS),and" Closed"(C). The Generic items were referred to DSI/RSB for consideration for incorporation into the next periodic update of the W STS in accordance with the provisions of NRR Office Letter No. 38. The PS Items were to be forwarded to the licensee (Ref. 5), and upon their response (Ref. 6), DL was to work with appropriate branches to achieve their resolution. The generic items ultimately arising out of the review are identified in the attachment. The original generic conclusions of DL are marked as G. Subsequent review by the writer and B. Sheron, Chief, DSI/RSB, of the original dispositions by DL, resulted in a transfer of a number of the items from the C and PS categories to the Generic category, including a new Multiplant 8903240078BBO3M69e ADOCK O PNV - PDR P

                     .                                                    Edward Butcher                                            .

Action (MPA)sub-category. TheseareidentifiedasG(RSB)andMPA(RSB) respectively. A number of additional Generic items arise from the joint response to the PS concerns by the licensee and Westinghouse under Reference 6, and these are marked as G (W). Robert B. A. Licciardo, Reactor Engineer (fluclear) Plant Systems Branch Division of Engineering and Systems Technology Attachments: , As stated l cc: T. Murley J. Sniezek I ( I

1 i i I

l l l

s, - q d 4 List of References l

1) Memorandum for Brian W. Sheron, Chief, Reactor Systems Branch, Division of Systems Integration, from Robert B. A. Licciardo, i Nuclear Engineer,

Subject:

" Review of McGuire Technical                                                                        l Specifications," dated June 11, 1984.

2) Memorandum for Darrell G. Eisenhut, from Robert M. Bernero,

             " Concerns on McGuire Technical Specifications," dated August 30, 1984.

3) Memo for Robert M. Bernero, Director, Division of Systems Integration from Hugh L. Thompson, Dircctor, Division of Licensing,

Subject:

! Disposition of Concerns Raised by R. Licciardo in His DP0 on the McGuire Technical Specification, dated May 28, 1985.

4) Letter from Nunzio J. Palladino, Chairman, USNRC, to the Honorable i Edward J. Markey, Chairman Subcommittee on Energy and Commerce, U.S. House of Representatives, dated May 17, 1985.

5) Letter to H. B. Tucker (Duke Power. Company) from Thomas M. Novak (DL),

Subject:

" Request for Comments on McGuire Technical Specification                                                             )

Concerns Resulting From Differing Professional Opinion," dated July 9, i 1985. l 6) Letter to H. R. Denton (NRC) from H. B. Tucker (DPCo) on

Subject:

             "NRC DP0 Concerns on McGuire Technical Specification," dated June 10, 1986.

l l i

                               .y                                                                     4 IDENTIFICATION OF GENERIC ITEMS CONFIRMED FOR WESTINGHOUSE STANDARD TECHNICAL SPECIFICATIONS                        1 BASED ON THE DETAILED REVIEW OF THE "PR0OF & REVIEW" COPY OF                                               l MCGUIRE UNITS l'& 2:    PROPOSED TECHNICAL SPECIFICATIONS l

PREPARED BY Robert B. A. Licciardo . l Reactor Engineer (Nuclear) United States Nuclear Reculatory Commission 1 Date: March 9, 1989 The generic items are identified by marginal marking in this attachment which is a copy of the criainal DP0 " Review of McGuire Technical Specifications" of June 11, 1984. The original generic conclusions of the Division of Licensing (DL)aremarkedasG. Subsequent review by the writer and B. Sheron, Chief, Division of Systems Integration / Reactor Systems Branch (DSI/RSB), of the original dispositions by DL, resulted in a transfer of a number of these items (from the closed (C) and Plant Specific (PS) categories) to the Gereric category, including a new Multi These are identified as .} G (RSB) and MPA (RSB) plant Action respectively. (MPA) sub-category. A number of additional Generic items arise from the joint response to the PS concerns by the licensee and Westinghouse under Reference 6, and these are marked as G (W).

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UNTTED STATES

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l MEMDRANDUM FOR: Brian W. Sheron,-Chief l Reacter Systems Branch B

                                                                   . Division of Systems Integration                                                                                                                              )

3 J FROM: Robert B. A. Licciardo ! Nuclear Engineer Reactor Systems Branch Division of Systems Integration SUE.1ECT: REVIEW DF .MCGUIRE TECHNICAL SPECIFICATIONS

REFERENCE:

a) Memo from Harold R. Denton, Director l Office : of Nuclear Reactor Regulation for Darrell G. Eisenhut, Director . Division of Licensing and d Roger J. Mattson, Director , Division of Systems Integration on the

Subject:

DIFFERING PROFESSIONAL , I DPINION OF MR LICCIARDD REGARDING MCGUIRI $ TECHNICAL SPECIFICATION and dated: March 21, 1981. I l b) Memo from Brian W, Sheron, Chief, RSE, DSI to 1 l Robert Liteiardo RSE, DSI dated April 11 1984 j

                                                                           'on the

Subject:

MCGUIRE TECHNICAL SPECIFICATIONS ASSIGNMENT j j

s J

1 I reference your memo to reference b) requesting review of the McGuire Technical J Specifications to an acceptable f,ormat, in response-to the requirement of ' reference a) for a coordinated review of the concerns arising from the writer's earlier OPO. Please find attached copy of a document entitled "McGuire Units 1 & 2: Proposed Technical Specifications; Review of Proof and Review Copy," which is in response to your request. The review is composed of two sections. The first section is entitled " Pre

Review Information" which details the Basis, Purpose and Resources, Schedule, l Evaluation Method, Regulatory Requirements and Licensing Consequences of the Review. The second section contains the Detailed Review. Since the staff required this detailed review to be conducted without any formal, or substantive informal discussion, both within and without RSS, I . presume that it is to be used as a basis for'the coordination stated in Harold R. Denton's letter to reference a), namely that "The Division of Systems Integration, in coordination with DL, shall hav.e people tha't are knowledgeable about the technical subjects raised by Mr. Licciardo, the standard technical s' specifications, and the McGuire technical specifications l

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l b writer considers *.het such a coo-dinated review including cons .ructive critique i is an essential consequence of any such document. The writer also believes that such construction must-be developed on the basis of responsible written f and signed comment within the Regulatory Framework. The writer would be pleased to participate in this coordination as required. j The writer is aware that RSS staff has received copies of the writer's fnitial proposed memo to T. M. Novak from R. W. Houston on the subject of: " STAFF . REVIEW OF PROOF AND REVIEW COPY OF PROPOSED TECHNICAL SPECIFICATIONS FOR I MCGUIRE UNITS 1 & 2" cated 06/15/83, and through this action is pleased te have made an early contribution to recent reviews of Technical Specifications for Operating License Applications. ( i Further, the writer has been informed that the above referenced memo (of ! l 06/15/83) was also provided to Westinghouse (W) and notes two subsequent I developments of significance: i

1) In response to a question from M. Wigdor concerning "Vogtle," on " Cold Overpressure Mitigation", W has now recently submitted a Topical report !

entitled " Cold Overpressure Mitigating Systems," cated February 1984, for i review by NRC.

2) W has recently reviewed its position on Reactor Coolant System (RCS)

Operability requirements in MODE 3 and from this has determined the need for additional operable RCS pumps over those required in the W STS for j the case c1 " Uncontrolled Ecd t1uster Control Assembly Bank Withdrawal i From a Suberitical Condition." ! l Both of the above items 1) and 2) were the' subject of specific concern in the referenced memo proposed by the writer, and it is encouraging to note the early

response by W to those safety issues. 1 R. B. A. Licciardo

{ l

Attachment:

As stated DISTRIBUTION Central File cc: H.R. Denton RSB R/F -; R. Mattson RLicciardo R/F R. W. Houston w/ attachment RLicciardo OP0 File l N. Lauben w/ attachment RLicciardo l l l Ub1:Kho i otrie > . . . . . . . .g.g. g.g.g.g. .j............. ................... .. . . . . . . . ...................... ec > .................... . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . ..................... ..................... ........;...........

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MC_G_UIRE UNITS 1 & 2: PROPOSED TECHNICAL SPECIFICATIONS REVIEW OF " PROOF & REVIEW COPY" Prepared By B ROBERT B. A. LICCIARDO Nuclear Engineer

RSB/DSI/RSB Date: June 12,1984 s y c s 2:c . ite 4 _ __ __-__ -_ __-__ - -

l TABLE OF CONTENTS

PRE REVIEW INFORMATION

        .           BASIS OF REVIEW
                                                                                                                      -l PURPOSE 0F REVIEW.

SCHEDULE AND RESOURCES EVALUATION METHOD REGULATORY REQUIREMENTS LICENSING CONSEQUENCES OF REVIEW INVITATION FOR COMMENT DETAILED REVIEW ADDENDA: Later' Items For Consideration - List of References Table 1. Sections Reviewed By' Reactor Systems Branch Table 2. Te'chnical Spec'ification Pages Affect'ed APPENDIX A: Technical Specifications'- Selected Relevant Regulations I l l l ~ l-I i l i I iii Revision A l

m-

                              .                                                                     1 1

Y

INTRODUCTION i By letter'to reference 1), the licensee proposed Technical. Specifications for d McGuire Unit 2 which were to be an integral part of the Operating License. The Licensee also proposed that these same Technical Specifications include detailed references to Unit 1 in a manner which did not impede its effective i use for Unit 2 but which would enable its.use for Unit 1 at a later date. 1 The Licensee considered an ultimate position in which both McGuire Units 1 and 2, would use the same Technical Specifications, with marginal adaptations. The application of these Technical Specifications to Unit I was achieved by application for a proposed, and issuance of a subsequent, licensing amendment at a later date. The Proof and Review copy which has been reviewed by the writer comprises a Westinghouse Standard Technical Specification, Revision 4, which had been marked up by the Licensee as a proposal for Units 2 (and 1). .This mark up was further reviewed by SSP 8 for conformance to the Westinghouse Standard Technical Specifications, and, by mutual agreement between the Licensee, ; NRR/DL and SSPB, subsequent changes had been made. This subsequent document' ! presented to RSB for review, contained no record of, or, safety evaluation - reports on, these changes which had been made including any relationship to l the then existing McGuire Unit 1 Technical Specification and the Final Safety Analysis Reports, or the Safety Evaluation Rtports, for McGuire Units 1 & 2. The writer has conducted the RSB portion of the review by a more detailed examination of those sections and related systems which are its primary responsibility as defined by the Standard Review Plan. These sections have been reviewed against the information in the Final Safety Analysis Report, the related Safety Evaluation Reports and additional information, as contained-in references 1 through 29. The Tableitems

2. reviewed are listed in Table 1 and the pages affected are listed in, i

l 06/01/84 iii Revision A

PRE REVIEW INFORMATION Basis of Review The starting basis for this review was the proposed memo to T. M. Novak from R. W. Houston dated 6/15/83, on the subject of: " Draft Review Of Prcof and Review Copy Of Proposed Technical Specifications For McGuire Units 1 & 2.' The Proof and Review Copy of the Proposed Technical Specifications For McGuire Units 1 and 2 from which the material for review by RS8 was extracted, was attached to a memo from C. O. Thomas (SSPB) to Brian W. Sheron (RSB) on the subject of " Proof and Review of McGuire - Units 1 and 2, Technical Specifications" and dated January 14, 1983. Purpose of Review and Resources The purpose of this review has been to enable a document which could be used to ser've the purpose of the request by Harold R. Denton in Reference a) namely:

             "The Divison of Systems Integration, in coordination with DL, shall have. people that are knowledgeable about the technical subjects raised by Mr. Licciardo, the standard technical specifications, and the McGuire technical specifications review the broad technical subjects and subgroups raised in the DPO."                                             '

For this purpose, RSB, asked the writer to identify the specific disparities of his concern,.and his basis'for them. Commencement of the task, as described under the section on " Schedule and Resources," disclosed more items of concern. To facilitate the preparation or a set of information within a time frame con- - sistent with the proposed purpose and schedule, the writer was asked by RSB to complete his task with minimal interchange both within and without RSB. This document presents the best evaluations by the writer under these conditions and must be considered as a starting basis for the follow-on coordinated review required from reference a). The writer wishes to acknowledge that during this review he has received the benefit of active discussions with ICSB personnel, namely T. G. Dunning, Section Leader, and F. Burrows, Reactor Engineer (Instr), on clarifying significant aspects of. Plant Instrumentation Logic. The responsibility for interpretation and conclusions in this document remains the writer's. Schedule ~ The starting basis for this review was the writer's proposed memo to T. M. Novak from R. W. Houston on the subject of Staff Review of Proof and Review Copy of the Proposed Technical Specifications for McGuire Units 1 & 2. By memo to reference a) dated March 21, 1984, Harold R. Denton required that:

     "The Division of Systems Integration, in coordination with OL, sha'l have people that are knowledgeable about the technical subjects raised by Mr. Licciardo,
    .the standard technical specifications, and the McGuire technical specifications 1                                         Revision A

__ _m _- _ - - - - _

review the broad technical subjects and subgroups raised in the DPO. As soon as the review approach is selected,'you are to provide me with a brief plan that describes how you plan to conduct the review, who is involved and your schedule ! for concluding the review. You should plan to document your review not later -l than July 1, 1984 or pro. tide a status report with a schedule by May 15, 1984." ' Commencing week ending March 31, 1984 the writer was asked by B. W. Sheron, Branch Chief, to develop a series of questions in accordance with his.later memo of April 11, 1984 for completion by April 27, 1984. ; On commencing this task, an audit was taken on other issues within the T.S. which had not received detailed attention because of relative priorities and ! the probabilities that because of the relatively simple nature of the related j operations, that the T.S. would be complete and accurate. This audit revealed ; that such was not the case and that relatively complex safety issues resided in l many locations of lesser perceived importance including footnotes, and descrip- i tions in the Basis, attached to the T,S. These concerns have renuired a near item by item check to ensure a maximum of surety. The schedule has been ex- ; tended on that basis but the need for closure has left a certain minimal area ! of unconfirmed concern. However, the above approach should now convince the licensee of his primary responsibility to ensure the accuracy and completeness of the Technical Speci-fications including a final detailed check and evaluation of no.t only the items that

are covered above, but residuals in the area of unconfirmed concern fpr RSB.

l Evaluation Method The evaluation has focused on the requirements of the process systems to meet ) Condition 1 Occurrences under normal operation in MODES 1 through 6. It has also focused on the capability of these same systems, and their protection systems [both Reactor Trip and Engineered Safeguards Features] to be available and to perform in accordance with acceptable calculated consequences of Condi-tion II, III and IV Occurrences, and other (Licensing Basis) events, as identified and evaluated in the Licensing Basis for MODES 1 through 6. 1 The term " evaluate," used throughout this review as e.g., in the phrase "The l licensee shall evaluate and propose" is to be interpreted as synonymous with the term " Safety Evaluation" as used in 10 CFR and includes the requirement to submit such an evaluation in response to related circumstances. - The term "proposo" is also synonymous with the term " propose" as used in 10 CFR 50.34(b)(6)(vi) " Proposed Technical Specifications prepared in accordance with the requirements of S50.36" and 10 CFR 650.59 " Changes, tests and experiments" in respect of " proposed change, test or experiment." Regulatory Requirements To facilitate ready reference, a set of " Selected Relevant Regulations" is provided in Appendix A, of which the following is a brief summary: 2 Revision A

   -t.
    .   '10 CFR 50.36        " Technical Specifications." This defines'the principal Require-ments which will be included in the Technical Specifications.

These include: 10 CFR 50.36(c)(1) " Safety limits,-limiting safety system settings and limiting controlLsettings."

10 CFR 50.36(c)(2) " Limiting conditions for operation" . 10 CFR 50.36(c)(3) " Surveillance requirements" 10 CFR 50.36(c)(4) " Design Features"' t ID CFR 50.36(c)(5) " Administrative controls"' ! 10 CFR 50.11 " Exceptions and= Exemptions from Licensing Requirements" l 10 CFR 50.12 " Specific Exemptions" These two Regulations define the basis.for granting exemptions from the requirenants of 10 CFR. -{ 10 CFR 50.34 " Contents of Applications: Technical Information" This provides the regulatory basis for a) Necessary descriptions of the facility and the need for i related Safety Evaluations for both'the PSAR and the FSAR. b) Within the PSAR, an identification and justification for the selection of those variables, conditions ~, or other items which are det. ermined as the result of preliminary safety i analysis and evaluation to be probable subjects of. technical- 3 specifications for the facility, with special' attention i given to those items which may'significantly influence the final' design. Reference 10.CFR 50.34,(a)(5). ' c) Within the FSAR, proposed technical specifications prepared in accordance with the requirements of $50.36. Reference l 10 CFR 50.34(b)(6)(vi) l 10 CFR 50.57 " Issuance of Operating License" l The particular relevant subsections are: 10 CFR 50.57(a)(1) - This ensures that the facility has been substantially constructed, in conformity with the construction permit and the application as amended. 10 CFR 50.57(a)(2) - which requires that "The facility will operate in conformity with the application as amended,..." l l l. 3 Revision A m- __ -.1._ . = _ _

                                                                                                                      . I J

4 10 CFR 50.57 (b) "Each operating license will include appro- j priate provisions with respect.to any uncompleted items of - construction and such limitations or conditions as are required .; to assure that operation during the period of the completion of I such items will not endanger public health and safety." i i 10 CFR 50.59 " Changes, Tests and Experiments" . 1 Sections of particular relevance are: 10 CFR 50.59(a)(1) - This permits changes from the FSAR providing they involve no change in the Technical Specification

do not involve an unreviewad safety question. 10 CFR 50.59(a)(2) - Defines an unreviewed safety question. 10 CFR 50.59(b) - Requires the licensee to keep a record of all changes made from the original FSAR and the related S'afety Evaluation, whether involvitg an unreviewed safety question or not. 10 CFR 50.59(c) provides that for these changes, tests and

                                        . experiments. involving an unreviewed safety question, the licensee shall submit an application for amendment of his license pursuant to 10 CFR 50.90.

10 CFR 50.90 " Application for amendment of license or construction permit" This pr.ovides that: "Whenever a holder of a license or construc-tion permit desires to amend the license or permit, application for an amendment shall be filed with the Commission, fully describing the changes desired, and following as far as appli-cable the form prescribed for original applications." 10 CFR 50.100 " Revocation, suspension, modification of licenses and construc-tion permits for cause." Licensing Consequences of Review - The consequences of the review in terms of the types of problems encountered in meeting regulatory requirements may be categorized as follows: J

1) Descriptions which are incomplete, ambiguous and errored, varying from relatively minor matters to matters of substantial importance to safety. )

l Except for relatively minor matters, this category has been considered 1 non conservative since they provide no sound basis for ensuring that the l detailed requirements of the Licensing Basis are specified for t'he l operating facility, j 1 1 4 Revision A , 4

Plant Engineering providing for unlimited operability of Process and 2) Protection Elements. Safety Evaluations have been submitted and accepted creating ~an element of the Licensing Basis [within the boundaries of unlimited operability]. The Technical Specifications are not in accordance.with the Licensing t3 asis by removing Operability Requirements without submitting necessary evaluations and proposals for evaluation by the NRC. For this situation, the general situation is that "The Licensee shall evaluate and propose." Examples include deletion of Operability Requirements for'RHR,. Component Cooling, RCS Loops, Elements of Reactor Trip System Instrumentation, and Engineered Safety Features Actuation System Instrumentation.

3) a) Plant Engineering with Operability Status limited by Plant Control or Protection Logic to certain MODES (and phases) of operation.

Safety Evaluations for the limited Operability Status have been sub-mitted and accepted as an element of the Licensing Basis. The Technical Specifications are not in accordance with the Licensing Basis Plant Protection Logic on which the safety was assessed e.g., ! Reactor Trip on ESFAS initiation in MODES 3 and 4 is not provided ' for in the Technical Specifications. The Licensee shall evaluate and propose.

3) b) Plant Engineering with Operability Status limited by' Plant Control Logic and related Safety Evaluations submitted. Review of submittals for Amendment may include an interfacing branch. SER issued contrary 4 to Regulations pertaining to that Branch. Examples include proposed j deletion of auto initiation of MD-AFW pumps below P-11 by manual 2

                -block, and deletioh of Pressurizer Water Level - High trip,                 f The p'roposed Technical Specification is in accordance with the Licensing Basis, but not in full accordance with Regulatory Require-ment. The licensee [should or] shall evaluate and propose.

This circumstance also introduces mixed and deficient protection , rationale for a large number of occurrences requiring protection aI under Regulatory Requirements. 1

'i

4) Plant Engineering with Operability limited by Plant Control Logic. q However, no Safety Evaluation has been submitted for the limited Opera-bility circumstances, which introduces unreviewed safety questions in the form of unforeseen and non-analyzed events. Examples include the absence of any " Low Flow" Reactor Trips below the P-7 permissive, and absence of many other Reactor Trips.

5 Revision A

The plant is inside the Licensing Basis Engineering which however has not been adequately evaluated. This is a situation in which Regulatory Requirements have not been met within the ensuing Licensing Basis since i an adequate clarification of and evaluation of the circumstances has not ! been undertaken. The licensee shall evaluate and propose.

5) The Safety Analysis Limits (in the form of response times) provided in the FSAR for ESFAs are in general less conservative than used in the evaluations of the Licensing Basis.

The Licensee shall evaluate and propose.

6) The response time provided may closely conform or agree to the Licensing Basis value, but the Licensing Basis value is contrary to Regulatory Requirements e.g., the Licensing Basis uses response times for AFW from non-safety related sources; whereas safety grade sources have a signifi-cantly greater response time. This delay may also impact response times for other ESFAs equipment.

The plant is inside the Licensing Basis Engineering which however has not been evaluated to Regulatory Requirements. The Licensee shall evaluate and propose.

7) a) Proposed Technical Specifications for major plant protection activi-ties which do not [ appear to] conform with the principal procedures '

described in the Licensing Basis. So that whilst the proposed Tech-nical Specifications are not in accordance and.also non-conservative, with respect to the Licensing Basis, they are also contrary to Regulatory Requirements. This applies particularly to 8 oration Control in MODES 1, 2, 3 and 4 and Emergency Core Cooling Systems in MODES 3, 4, and 5. No i evaluation and proposals are submitted. The Licensee shall evaluate and propose.

7) b) Also, as a result of 7)a), we have discussed possible modifications to these proposed Technical Specifications, which may make them -

acceptable providing appropriate projections are added and suitable evaluations proposed. Examples include the virtual absence of any necessary protection (including constraints) to ensure RCS safety to Regulatory Require-ments under Condition II, III and IV occurrences in MODES 3, 4 and 5 due in part to the Boration Control disparity mentioned in 7 a) above.

8) The absence of necessary correlations between surveillance requirements for equipment performance and that performance necessary to achieve the required Plant Protection under Condition II, III and IV Occurrences.

l 6 Revision A

l l An example includes Aux FW distribution to remaining intact Steam Generators in 'a Main Feed Line Rupture Event in which two Steam Generators providing steam to the Turbine Driven AFW- Pump are ultimstely faulted. The licensee shall evaluate and propose.

9) It is a fact that engineering and construction of a nuclear facility must be checked on an element by element basis to ensure that the eno'nnity of all the interfaces meet as required to enable final assembly and startup.

Similarly, with Technical Specifications, unless they are likewise checked on an element by element basis, there will be no guarantee that the plant will have the level of safety proposed in the Licensing Basis Docments. The Licensee has primary responsibility for this element by element check and our review together with responses from the requested evaluations and proposals will . reflect the consequences of the exercise of that responsibility. l Invitation For Comment The writer would welcome written and signed comments within the Regulatory Framework, on this Review. Re f e rene es a) Memo 4 rom Harold R. Dentone Director Office of Nuclear Reactor Regulation f or Darre t t G. Eisenhut, Director Division of' Licensing and Roger J. Matt sone Director Division of Systems Integration on the

Subject:

DIFFERING PROFESSIONAL OPINION OF MR. LICCIARDO REGARDING MCGUIRE TECHNICAL SPECIFICATION and dated: March 21, 1984 b) Memo from Brian W. Sherone Chieff RSB, DSI to Robert Licciardo RSBr DSI dated April 11r 1984 on the

Subject:

MCGUIRE TECHNICAL SPECIFICATIONS ASSIGNMENT 4 6 7 Revision A

1 4'

i

        .                                                                                    3                        1
                                                                                                                ,1 l

MCG IRE UNITS 1 & 2: PROPOSED TECHNICAL SPECIFICATIONS DETAILED REVIEW OF " PROOF ~& REVIEW" COPY i PREPARED BY Robert B. A. Licciardo )

                                                      ' Nuclear Engineer.

RSB/DSI/RSRS Date: June 12,1984 : , l l . l 1 e i .

i. 06/07/84 Revision A l

i

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

{ APPLICABILITY: M0025 1 and 2. ACTION: Whenever the point defined by the combination of the highest operating loop aver' age temperature and THERMAL POWER has exceeded the appropriate pressurizer pressure line, be in HOT STANDBY within 1 hour, 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, and the values in this figure. Reference 7, Figure 15.1.1-1, revision 7 is the existing licensing basis; it provides different ordinates, T avg vs AT and includes descriptions of related acceptance criteria and limits which l 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 loop (individual) compared with AT loop set point (individual), in the reactor protection system. Any such representation should also provide the basis for the SET-POINT methodology for each unit including values of all the parameters necessary to calculate OVERTEMPERATURE AT and OVERPOWER AT. SET POINTS of related Table 2.2-1, REACTOR TRIP SYSTEM INSTRUMENT TRIP SET POINTS; this will ensure a complete set of Licensing Basis data against which the 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 inappropriate. Neutron flux limits and related action statements are addressed under T.S. Sec-tion 3.4, [ Nuclear] Power Distribution Limits. d) References to three loop operation should be deleted as the plant is not licensed for such operation. l 06/01/84 1 Revision A

I e) Concerning description under Section 2.1.1 above. We propose this de- I scription should clarify that the " combinations" presented are those allowed - under " Anticipated Operational Occurrences" and not steady state conditions. j k f) The FSAR does describe a constrained set of thermal hydraulic parameters -

for the Reactor Coolant System under steady state normal operating con- ] ditions upon.which " plant safety" under Condition II, III and IV Occur- i rences is established. These are generally described in reference 7, under Section 15.1.2, Table 15.l.2-2, and the pr'ogrammed T 3yg 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 i 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 steady state set points would be represented by a single line showing programmed 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 Licensing 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. i ACTION: 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 within - 1 hour, and comply with the requirements of Specification 6.7.1. MODES 3, 4 and 5 Whenever the Reactor Coolant System pressure has exceeded 2735 psig, reduce the Reactor Coolant System pressure to within its limit within 5 minutes, and comply with the requirements of Specification 6.7.1." EVALUATION a) Is there not a need to forewarn the operator that as for 2.1.1, for normal steady state operation, the RCS pressurizer pressure shall not exceed the 06/01/84 2 Revision A I i

l i 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 from the existing Licensing Basis and an appropriate request for a T.S. change !

      .          would be necessary.

b) As for Section' 2.1.1 above, is _it not appropriate to clarify that the RCS Coolant System pressure shall not exceed [2735] psig under any Anticipated Operational Occurrence or Design Basis Accident. c) Where.in the RCS system is the pressure limit to be observed eg 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 added to the calculated pressurizer pressure." What provision has been made in the specified value or related instrumentation to conservatively I account for this necessary correction. d) Please clarify that the value of 2735 psig is an actual Safety Limit, being 110% of the Design Pressure of 2485 psig (reference 3, Table 5.2.2-2) i and how is such a value determined by the operator when no set point, ' allowable values and channel errors are provided for or defined. j e) Concerning Action Statement: MODES 1 & 2. This should consider restora-tion of the RCS pressure to its required value for steady state operation rather than within the 2735 psig limit. Should MODE 3 also be included in the action statement for MODES 1 & 2 as j generally identical concerns prevail except for the limited Applicability )

              ,  of Appendix G in T.S. Figs. 3.4-2.                                              .

f) Concerning MODES 3, 4 & 5. i l How is the pressure limit of 2735 psig applicable to MODES 4 and 5 when j reduced RCS temps. will cause consideration of constrained Pressure / ' Temperature limits [to- Appendix G requirements] in T.S. Section 3/4.4.9. Further, even MODE 3 has an ppendix G limits of <2500 psig at RCS temps. of <350.F; reference T.S. Figs. 3.4-2.

SUMMARY

The current representation of Safety Limits for RCS pressure in this Sec- _ tion 2.1.2 is non-conservertive with respect to the Licensing Basis. The

        . Licensee shall evaluate and propose.

06/01/84 3 Revision A

                                                                                                             .R
    .             TABLE 2.2-1.      REACTOR TRIP INSTRUMENTATION SET POINTS                                        J These have been checked against reference 18, Westinghouse (W) RPS/ESFAS Set Point Methodology, Table 3-4 and NOTE:FOR TABLE 3-4 on page 3-13, which is               '

q described as applicable to McGuire Unit 1, 50-369. At this date, the assump- i I tion has been made that this information also applies to McGuire Unit 2, Docket No. 50-370. Please docket this fact or otherwise provide the alternate information. The writer finds the general approach to representing Trip Setpoints as g or 1 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.

Item 3. Power Rate, Neutron Flux, High Positive Rate Will a time constant of >2 seconds result in a slower response time, which is less conservative. 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? 1 Reference 18 page 3-13, concerning Set Point Methodology adv.ises 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 Licensee shall evaluate and propose                                                              )

Item 5: TS incomplete; should read as: Intermediate Range, [High] neutron flux. I Item 9: Pressurizer Pressure-Low , The specified Trip Setpoint & Allowable values agree with those provided under setpoint methodology in reference 18. A disparity does exist between the related SAFETY ANALYSIS LIMITS given as used in Safety Analysis, i.e,1845 psig in SETPOINT METHODOLOGY / Reference 18, Table 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 identify the correct value. [ Note also disparity with reference 7, " Analysis of Inadvertent Operation of ECCS Ouring 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 respect to the other values used above.] The Licensee shall review and clarify. Item 17: The existing descriptor " Safety Injection Input from ESF" should be l replaced by " Reactor Trip from ESFAS." 1 06/01/84 4 Revision A

-v I 1 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 Pressurizer Pressure-Low 17b) Containment Pressure-High 17c) Low Steam Line Pressure (subject to P-11 block) 17a) Manual Safety Injection Item 12: Low Reactor Coolant Flow

a. Concerning Reactor Trip on " Low-Reactor Coolant Flow 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 ie under frequency, undervoltage, loss of RCP breaker. Such anticipatory signals are provided below P-8 when safety consequences are more conservative for this facility. (See later 12b.) Is this adequate conformance to diversify requi,re-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 operation with only 3 loops operating in MODES 1 and 2 below P-8. Please explain why you therefore propose a trip based on Loss of Flow in 2 loops instead of only one, at these conditions and which is not in conformance with GDC 20, " Protection System Functions." Information is provided 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 basis and should therefore be excluded. This licensee should evaluate our concerns in items 12a and 12b above in conjunction with those of item 18.b.a of this same review of Table 2.2-1,.and propose. This can be interpreted as a generic issue. Item 13: Concerning Steam Generator Level-Low, Low Reference 18, page 3-13 Note 12 describes the Safety Analysis Limit for this item as the value in Table 2.2-1 of the W STS plus 10%. For conservatism, should the Safety Analysis Limit 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 (10% power), Reactor Trip deriving from undervoltage of the Reactor Coolant Pumps is also initiated, reference 7 Section 15.2.5.1 and reference 5, figure 7.2.1-1 l i 06/01/84 5 Revision A

i e

note 4. It is proposed that a notation to-this effect should appear under this item. - l Item 21 (Proposed): [ 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 undervoltage trip relay above Permissive 7 (10% power); reference 7, section 15.2.5.1. Item 16b: Low Power Reactor Trips Block, P-7 . a) This T.S. provides that when power level is less then Permissive P7 (with I 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 FSAR in reference 5, item 7.2.2.1.2d which describes this permissive , provides no safety evaluation of the consequences. Accident Analysis in 1 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 acceptable RCS response under these particular circumstance. The existing FSAR, reference 7, Section 15.2.9.2 and related Table 15.2.9-1 shows acceptable 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 additional events potentially arising from this item which have not , been analyzed. These include a circumstance in which a normal turbine 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 Turbine Power signals are reduced below 10% (P-7) so that reactor trip on this loss of power event could 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, j that below P-7, a number of other reactor trips are also blocked including Pres- l surizer Water Level-High, Pressurizer Pressure-Low and Pressurizer Pressure-High] ' The situation is one in which Condition II, III and IV occurrences are not protected in accordance with GOC 20, Protection System Functions: "The . protretion system shall be designed (1) to initiate automatically the operation of appropriate systems including the reactivity control systems, to assure that specified acceptable fuel design limits are not exceeded as 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 combination with the initiating occur-rence has not been evaluated. It has not been Regulatory Practice to allow a Condition II occurrence to be followed by a Condition III or IV occurrence in the course of protective actions. 06/01/84 6 Revision A 1

1 i The licensee should evaluate the restoration of reactor trip on " low flow" trips I down to and including MODE 2 (MODES 3-5 are discussed later) to be in conformance I with G.D.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 3 Operating Occurrences and Postulated Accidents are currently satisfied, making i appropriate allowances for any manual Operator Action required. These events 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 l Level-High and Pressurizer Pressure - Low 3re 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-able, but does not trip the Reactor at these low power conditions (below P-8).] Until the required re-evaluation is completed, the proposed T.S. must be considered non-conservative in respect to Regulatory Requirements. Additionally 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 provides that when power level is less than Permissive P7 (with P10 (Nuclear) or P13 (Turbine) powers of less than 10%), reactor trip on !

Pressurizer Pressure-Low and Pressurizer Water Level-High are both blocked.

                   .c(i) Concerning Block of Pressurizer Pressure Low - Reactor Trip:

The FSAR in reference 5, item 7.2.1.1.2.C.1 states that this trip is not required at low power levels. The pressurizer pressure low - reactor trips are used as both primary and back up in a number of Condition II Condition III and Condition IV occurrences, all involving breaks in the primary and secondary systems, reference 7, table 7.2.1-4 (3 of 5). Although safety injection is subsequently employed in almost all these situations, earlier reactor trip on pressurizer pressure low - is depended __ upon instead of the later reactor trip on pressurizer pressure low - (Safety Injection). The worst situation for most of these accidents is that of maximum power level reference 7, Table 15.1.2-2. No evaluations are provided for zero power level. It is possible for these breaks in the primary and secondary systems to occur at less than 10% power level down to and including the startup condition (with 4 RCS loops running) ie MODES 1 & 2. (Such breaks in MODES 3-5 are discussed later). With the proposed TS, reactor trips for these breaks would be delayed to be initiated later by the ESFAS (SI) related signals. The licensee should provide a safety evalution of these circumstances and which is not based upon arguments relating to probability of the events. The evaluation should provide 06/01/84 7 Revision A

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. ' Until there has been a re-evaluation of these circumstances, the proposed T.S. - must be considered non-conservative in respect to Regulatory Requirements. Additionally it can be interpreted as a Generic Issue. 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 respect to the absence of the signal caused by P7. cii) Concerning Block of Pressurizer 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 Assembly 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 Turbine 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 dump) with the expectancy of no values exceeding normal control band [and thereby not approaching T.S. Limits). The blockage of these trips is consistent with the Design Basis Events _and ex- , pected behavior of the Control System. However this does not address the fact j that Design Basis events only define the outer envelope of expected severity which is expected to cover a large number of less severe occurrences, undefined. ; It appears singularly inappropriate to remove these protection devices which k could play a primary or backup role in such circumstances. For example, refer-ence 5, page 72-27 item 7.2.2.3.4, " Pressurizer Water Level," describes the role of the Pressure Water Level trip in preventing liquid Coolant discharge through the safety valves during a failure of the Pressurizer Water Level (PWL) controller _ at full power. Failure of PWL controller could fill the pressurizer within hour or longer, but T.S. Table 4.3-1 shows a channel check on only a shift 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 from 2 out of 3 so that no alert is given the operator which would be contrary to the requirement of the FSAR. There is no discussion on the importance of its use at low powers although the general System Description provided under Section 7.2.1.1 and its 06/01/84 8 Revision A

     .       protective actions is no less appropriate at 0-10% power, as it is at higher power levels.

It is proposed, reference 5 page 7.2-6 that Pressurizer Water Level-High 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 pressurizer in MODE 4 and the Pressurizer Level Control can be placed in AUTO. Considering the attention required of all other manual actions during MODES 4 through 2, it is not appropriate 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 Pressurizer 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. ; This item can be interpreted as a generic issue. This could be considered non-conservative in respect to Regulatory Requirements because of the absence of l l automatic protection in accordance with 10 CFR 50, GDC 20 " Protection System Functions," both for reactivity control systems, and overpressure protection systems. c(iii) The absence of permissive P-7 [on P-10 and P-13] introduces new events to ' 1 evaluate for safety. This requires related Safety Analyses Limits and 4 the Licensee shall advise what these are for each of P-10 and P-13 and ' how these are combined for P-7. . Item 18(f). Proposed new item: High Power Reactor Trip on Turbine Trip; Black 1 by absence of P-8. { { 1 The Anticipatory Reactor Trip on Turbine Trip required by TMI Action Plan II.K.3.12, is bypassed below P-8. The SER is provided in reference 15, l Item II.K.3.12, and reference 21 for McGuire Unit 1. We have issued no related final SER for McGuire 2 at this time. Note the related Basis will need to be amended. 1 Item: Loss of " POWER" l Their is a need to prescribe the conditions under which a reactor would trip directly from a " Loss of Power" condition other than those deriving from other Functional Units. This is a substantial omission from t'he' Tech- - nical Specifications. Item: General - This is a need to identify potential blockage of each of these Reactor Trip Functions by Plant logic and any related manual action, e.g., 3 P-7, j P-11 with manual blockage etc. This enables improved perception of real levels of engineered protection than is currently available. Table 3.3-1 contains only approximate information concerning plant situations at which protection levels are changed. It also contains NON-0PERABILITY MODES which are not pre-determined by Plant logic. 06/01/84 9 Revision A i

          ._    __.-_.__--______-__.-_m.__.___._______________-_--_-_._m._                  _ _ - - _ .

i

 .       SECTION 3.4.1     REACTIVITY CONTROL SYSTEMS Section 3/4.1.1 BORATION CONTROL / APPLICABLE MODES 1, 2*, 3 and 4.
       ~                                                                                                                                                      .

T.S. Pages 3/4 1-1, 2, 2a: Reference 16; page Q 212-47e states " Operating

b. Instructions require that boron concentration be increased to at least the cold

, (gg) shutdown boron concentration before cooldown is initiated. This requirement insures a minimum of 1% delta k/k shutdown margin at an RCS temperature of 3 200 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 I MODE 3 down to and including MODE 5, please advise why this does not appear in the Technical 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 K/K over l MODES 1 through 4 should be modified to exclude MODES 3 and 4, and SHUTDOWN j MARGIN T should be changed from >200 F to 1557 . AVG l A new T.S. Page 3/4 1-2(a) should be added for BORATION CONTROL SYSTEMS in l MODES 3 through 5, from T < 557 F through 140 F, providing that the boron , l o concentration in the RCS $h211 be increased to a value which will give a i shutdown margin of 1% delta K/K at 200 F. 1 Safety Significance: These actions are necessary to bring the safety status 1 of the plant into conformance with the Licensing Basis. Without this, the i plant is in a less than conservative MODE which has not been evaluated. I Further, it appears that OPERABILITY REQUIREMENTS of Table 3.3-1, REACTOR TRIP j SYSTEM INSTRUMENTATION and TABLE 3.3-3 ESFAS INSTRUMENTATION may be conditioned on these higher Boron Concentrations so that ommission.of Additional Boron 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, g < 557 F. The proposed T.S. might be acceptable if all events were analyzed in MODES 3 ! through 5 and the OPERABILITY REQUIREMENTS OF TABLES 3.3-1 and 3.3-3 reviewed. l Reference 11, page 15-2, first para, precludes any boron dilution after a reactor scram until the neutron flux level is.below the level of the source I range high flux level alarm. This is effectively an LCO that is not included l^ 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 3/4.1.1 and propose. TS Page 3/4 1-6. MINIMUM TEMPERATURE FOR CRITICALITY The existing minimum temperature for criticality (in MODES 1 and 2) is given as 551 F. Please advise why this value is less than the programmed set point minimum value of 557 F in reference 20, fig. 5.3.3-1. Accident evaluations for events from zero power are predicated upon this set point of 557 , and any 06/01/84 10 Revision A _ . . - _ _ _ _ . - . . _ _ . _ _ _ _ _ _ _ _ _ _ ..___-..-___.m___-___.-_____.____.___m ___

4

       -                                                                                                                        i I

variation therefrom in either direction would be unacceptable. Reference our comments under Section 2.1.1.f. , An example of a safety impact is for the Design Basis Main Steam Line Break Event which is initiated from 2ero power in MODE 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 TS Section 3.1.1.4 should therefore I 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 MODES 1 and 2. ) The proposed T.S. is non-conservative with respect to the Licensing Basis. The j Licensee shall evaluate, including the above concerns, and propose. ) i Section 3/4.1.2 BORATION SYSTEMS T.S. Page 3/4 1-7: Concerning "B0 RATION SYSTEM, FLOW PATH - SHUTDOWN. APPLICABLE MODES 5 and 6: The current T.S. requires an (unidentified) charging pump to supply Boron to l the RCS. Current Licensing constraints on ECCS operation discussed under i Section 3/4.5 Emergency core cooling systems" require that only one centrifugal 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 MODE 4, a similar )

and parallel requirement for overpressure protection in-the RHR mode with ) water solid operation extends this requirement through 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 Appendix G Limit is not reached. , 1 l Charging pump requirements in MODE 6 are defined by reference 10, Sec- ) tion 15 2.4.2, item 3 under " Dilution-During Refueling" in which a pre- l condition for the " uncontrolled Boron Dilution Event" is that "the charging pumps are inopei d ive." These circumstance permit only one charging pump, which must be a centrifugal pump only, in operation from " standby (at 1000 psig/425 F) through to MODE 5"; therefore the term SHUT 00WN in the title and the APPLICABLE MODES 5 and 6

      .                          should be replaced by these conditions. Also, the description of the charging                -

pump should be expanded by the term " centrifugal" together with the proviso that "this centrifugal charging pump also be the same and only pump allowed for i ECCS and other operations under these circumstances." l l The proposed T.S. is non-conservative in respect of the Licensing Basis. The Licensee shall evaluate and propose. t l 06/01/84 11 Revision A

                                                                                                                       .          1 i

T.S. Page 3/4 1-8. Concerning: " FLOW PATHS - OPERATING" in APPLICABLE MODES 1 l

2, 3 and 4. j The Licensing Basis ECCS requirements discussed under Section 3/4.5 EMERGENCY - CORE COOLING SYSTEMS of this report do not constrain charging pump operation above 1000 psig/425 F. Therefore the existing provisions on this T.S. page for charging pumps remain valid with the exception that APPLICABLE MODE 4 should be deleted and MODE 3 must be conditioned as MODE 3 (Down to 1000 psig/425 F). Further the title should be changed to incorporate these constraints. ! 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 Boration ; Requirements adopted out of item "Section 3/4.1.1" of this report. l l T.S. Page 3/4 1-9 concerning: CHARGING PUMP-SHUT 00WN i Consistent with the work of the previous TS Section 3/4 1-7 of this report, this title snould be changed to: CHARGING PUMP "Standbye (at 1000 psig/ 425 F) through to MODE 5. Additionally, under subsection 3.1.2.3 modify to only one centrifugal charging pump shall be OPERABLE. APPLICABILITY is changed

   - from Ff66ES 5 and 6 to MODE 3 (at < 1000 psig/425 F), 4 and 5. MODE 6 is deleted.

Surveillance Requii2ments under subsection 4.1.2.3.2 must reflect the require-ments of later SECTION 3/4.5 ECCS of this report in which "All centrifugal, [and reciprocating] charging pumps excluding the required OPERABLE pump shall be demonstrated inoperable by" additional features to those already described in this subsection, namely, "by verifying that the motor circuit breakers are secure'd in the open position by being opened, locked and tagged; the alternate of isolation from the Reactor Coolant System by at least two isolation valves ! with breakers for the valve operators being open, locked and tagged has not been provided. (reference 12, page 6-6 concerning racking and locking out of pumps; also reference 11, pages Q212-47 and 47a) The proposed T.S. is non-conservative with respect to the Licensing Basis. The Licensee shall evaluate and propose. j MODES 1 2 n 4 - 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 MODE 4, and MODE 3 co'nditioned to (down to 1000 psig/425 F) Item 4.1.2.4.2 under SURVEILLANCE REQUIREMENTS does not now apply since it refers to conditions 5 300 F which are not now covered by this section, being limited to a minimum of 1000 psig/425 F in' MODE 3. The same comment applies to footnote #_ concerning one only centrifugal charging pump at 5, 300 F. The proposed T.S. is non-conservative with respect to the Licensing Basis. The Licensee shall evaluate and propose 06/01/84 12 Revision A

m . . . . .. . . . . - - . . . . . . . . . 1 i

           ,T_._S. Page 3/4 1-11 Concerning:                 B0 RATED WATER SOURCE - SHUTDOWN This title (and related Applicability MODES 5 and 6) should be changed to BORATED WATER SOURCE - MODE 3 (1000 psig/425 F) THROUGH TO MODE 5, to be compatible with the changed title to TS pages.                     3/4 1-7 and 3/4 1-9 discussed earlier since this page refers to borated water sources for situations there                                           j described.

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 l be included under 4.1.2.5.a.4) in which the borated water source would be demon- l strated OPERABLE by verifying minimium levels in the system. Further, an additional surveillance should verify the availability of Level Detection (2 indicators / tank) and related high, low and low-low level' alarms. Clarify whether the LCD values proposed are Safety Analysis Limits or Set Point Values. j An appropriate modification may need to be made to the Boron Concentrations and , volumetric requirements in the Boric Acid Storage System in these MODES 3 (1000 psig/425 ) through 5 to provide for the increased Boron Concentrations required from the Licensing Basis in these MODES discussed in this report under T.S. page 3/4 1-1, 2 and 2a. - Why is the refueling water storage in MODE 5 proposed as only 26,000 gallons when reference 8, page Q212-57, revision 25, under Case-3 provides that in MODE 5, in the event of loss of cooling by a fail closed RHR/RCS isolation valve the charging pump could provide feed and bleed cooling through the PORVs for up to 5 hours from the RWST and subsequently the RHR pump and heat exchanger would re-circulate and cool from the containment sump. Would not this require an unchanged requirement froni MODES 1 through 4 of at least 372,100 gallons. The proposed T.S is non-conservative in respect to the Licensing Basis. The Licensee shall evaluate, including all our concerns above under T.S. Page 3/4 1-11, and propose. T.S. Page 3/4 1-12 concerning: BORATED WATER SOURCES - OPERATING (in related Applicable MODES 1, 2, 3 and 4) 1 This title, and related applicability modes, should be changed to: B0 RATED WATER 50JRCES - MODES 1, 2, and 3 (Down to 1000 psig/425 F) to be compatible with the changed title to T.S. Pages 3/4 1-8 and 3/4 1-10 discussed earlier, since this page refers to borated water sources for the situations there described. Additionally, [by letter to reference 17] the Licensee did commit 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 3.1.2.6.a.4). Additional surveillance requirements l I should be included under 4.1.2.6.a.4) in which the borated water source would be demonstrated OPERABLE by verifying minimum levels in the system. I J 06/01/84 13 Revision A )

1

m i Further, an additional surveillance should verify the availability of Level Detection (2 indicaters/ tank) and related high, low and low-low level alarms. , Clarify whether the LC0 values given are Safety Analysis Limits or Set Point Limits. i An appropriate modification may need to be made to the Boron Concentrations and volumetric requirements in the Boric 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 TS page 3/4 1-1, 2 and 2a. The absence of required LCOs makes the proposed T.S. less conservative than the Licensing Basis. The Licensee shall evaluate, including our concerns under TS Pages 3/4 1-12, and propose. T.S Page 3/4 1-13a. Proposed concerning: INSTRUMENTATION IN MODES 3, 4, 5 and 6 SER Supp 1, reference 11 page 15-2 requires a Technical Specification that "During startup and shutdown, the applicant 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 level of 5 times the background level. The licensee is to maintain the source range alarm setpoint at this l level or lower any tifne the plant is in the cold shutdown Mode. The set i point is to be checked and adjusted on a weekly basis if in the cold shutdown l mode for an extended period." This SER requirement has not been provided in the Technical Specifications. l Please discuss provision under a proposed new item under Section 3/4.1 l REACTIVITY CONTROL SYSTEMS, entitled " INSTRUMENTATION" in which these require-l 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 applicable only to MODE 6. Since it is a part of " Reactivity Control Systems" and applicable over additional MODES, it should be provided in this i context also as discussed above. l The proposed T.S. is less conservative than the Licensing Basis. The Licensee shall evaluate and propose. T.S. Page 3/4 1-20 Concerning: SHUT 00WN R00 INSERTION LIMITS T.S. Page 3/4 1-21 Concerning: CONTROL R00 INSERTION LIMITS a) Specifications for limiting conditions of operation on the positions of these movable control assemblies apply only to MDDES 1 & 2. There is no Technical specification on positions in MODES 3-5 although T.S. Page 3/4 1-18 concerning " Position Indication system - shutdown" requires operability of a Rod Position indication system in MODES 3 through 5'when the reactor trip system breakers are in the closed position. 1 06/01/84 14 Revision A

i It is proposed that in general, Technical specifications are required by 10 CFR 30:46 to be placed on the limits of movable control assemblies in these modes to limit the consequenceslof Condition II, III and IV events which may occur, unless analyses and evaluations show that these are unnecessary. An example of the need is reflected in the memo to reference 26 in which rod positions for Boron Dilution events are specified from Refueling through to Hot standby as All Rods Out (Mode 6, Refueling) and, All Rods In with Most 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. The Boron Dilution event for McGuire has "apparently been" made acceptable by procedures requiring the RCS to be filled with Borated (approx 2000 ppm) water from the refueling water storage tank prior to " Start Up"; reference 7, page 15.2-15, revision 10. Reference earlier discussion on TS. Pages 3/41-1, 2 and 2 a. This is an LCO and should appear in the proposed T.S. l With the existing T.S. without the required increase.in Boron 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 accepta'le o safety evaluation , is submitted to show the contrary. ] In general, also, the same concern applies to any other Condition II,.III and IV occurrence which can lead to a return to power in these Modes. Until these i circumstances can be shown to result in acceptable 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 ' l given to the reduced operability requirements for all Reactor Trip Instrumen- ) tation and Engineered Safety Features Actuation Instrumentation in these ] MODES (3 through 5). This is particularly significant with the proposed T.S. l on Boration Control where resulting shutdown margins are substantially less than these provided by the current Licensing Basis. , The Licensee shall provide analyses and related safety evaluations to justify his current absence of Technical Specifications in respect of SHUTOOWN and CONTROL R00 positions during MODES 3 through 5. Without this, the proposed T.S are non-conservative with respect to the Licensing Basis. b) Overpower (AT) and overtemperature (AT) protection systems incorporate

 .       automatic limits (Rod stops) on control rod insertion to maintain Safety                  -

Analysis Limits on " Power Distribution" in the Reactor Core during power runback. Please advise why there are no surveillance limits and requirements for these Rod stops in your Technical Specifications to mcet the requirements of 10 CFR 50.36. Without these, the proposed T.S. must be considered non-conservative.

l l 06/01/84 15 Revision A l l l

Section 3/4.2 POWER DISTRIBUTION LIMITS - l Section 3/4.2.1 THROUGH 3/4.2.4 POWER DISTRIBUTION LIMITS i RSB has not reviewed these sections on the understanding that they are the primary responsibility of Core Performance Branch. Section 3/4.2.5 DNB PARAMETERS AND TABLE 3.2-1 ONB PARAMETERS The , current information does not adequately represent all those perameters necessary to ensure " acceptable" RCS operations, including DNB, under all Licensing Basis Conditions II, III and IV. The necessary parameters are discussed and described under Section 2.1.1 Reactor Core, item f, of this report. If they are logically represented under 2.1.1. [and elsewhere), why are they also represented here? Evaluation a) ONB presents only one Acceptance Criteria for acceptable operation of the RCS: There are others including Fuel element clad failure and Appendix K , requirements depending upon the occurrence being considered. Additionally l there are RCS overpressure, steam generator overpressure and Hot Leg Boiling Criteria. As indicated in our comment in Section 2.1.1, item f, initial conditions which j

        ,              cover a larger N of variables than those presented in Table 3.2.1, in combina-tion, determine RCS safety in the necessarily broadest sense.

It is suggested that this section be deleted, and the relevant information be supplied under T.S. Sections 2.1.1 where it belongs and where it has been discussed. b) Concerning Table 3.2.1. The value for Reactor Coolant System T a given I as1593FisnotinaccordancewiththeFSAR, reference 3,FigureSgg3-1 i d[ where.a value of 588.1 F is given as the programmed T for RATED THERMAL J POWER Conditions. Pleaseexplainthedifferenceand8xhlainwhysetpointand allowable values should not be provided. As a Setpoint, the proposed TS value is non-conservative with respect to the Licensing Basis. . dy Please explain why a related power level has not been ascribed to this temperature. . f l gg Please explain why programmed T of 557.0 F (also reference 3, Figure 5.3.3-1 l hasnotbeengivenforzeropow3E9 operation (Reference again our Section 2.1.1 . l t .

                       . item f).
                       'c ) Concerning Table 3.2-1 Pressurizer Pressure. Please explain the basis for the given value of C 2230 psia when information in reference 20, Table 4.1-1 gg           (1 of 3) shows a " System Pressure, Nominal" of 2250 psia and Section 15.1.2.2,                              ,

Table 15.1.2-2 makes provision for a total of 30 psi for steady state fluctu-ations and measurement error. Have you quoted a Setpoint value, or an allowable 06/01/84 16 Revision A

value; both should be available. As a Setpoint, the proposed T.S. value is non-conservative with respect to the Licensing Basis for DNBR, and conservative for 3 I overpressure protection. d) Why should not programmed T,yg be provided under T.S. Section 2.1.1 , dh,gq 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 f) As discussed in Section 2.1.1, Subsection f, additional parameters necessary I 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 . e5

_ Levels under Section 3/4.4.5 T.S. Page 3/4 4-11). derd; CONCLUSION The parameters proposed by the T.S. as "DNBR PARAMETER" under TABLE 3.2-1 are an incomplete set and inadequately defined in terms of Set Points. Allowable -

Values and Safety Analysis limits. All this necessary information is available

                                                                                                                                      .q from the existing Licensing Basis and their incomplete and inadequate repre-                       j sentation creates a non-conservative situation with respect to the Licensing                       1 Basis. The Licensee shall evaluate and propose. This is only partly a generic problem arising from an inadequate representation in the W STS.
                                                                                                                         .              l I

ie l l l 06/01/84 17 Revision A

TABLE 3.3-1 REACTOR TRIP SYSTEM INSTRUMENTATION T.5. Page 3/4 3-2. Item 6c: Source Range, Neutron Flux Does this channel provide an alarm only function, o' r an alarm plus trip function. During shutdown in MODES 3, 4 and 5, with reactor trip system breakers open, r Source Range, Neutron Flux, channel operability requirements specify only one channel operable, and if this same channel is being used to meet the Boron dilution alarm requirements of proposed T.S. Page 3/41-13 (a), then it is not in accordance with the Boron Oilution Requirements of the FSAR.for which at least 2 operable channels would be required; reference 8, page Q212-24, item 212.58. The Lict osee shall evaluate and propose. Currently, this appears non-conserveve. ) Item 6a: This Technical Specification concerning Operability of the Source Range Neutron Flux is unclear. It species operability of the Source Range Neutron Fiux trip Below the P-6 (intermediate Range Neutron Flux Setpoint) during startup in MODE 2; the Licensee shall advise if this " start up" channel is required to be Operable to get Reactor. trin in MODES 3, 4 and 5.

Items 1 through 5: The FSAR, Refe'rence 5, ib le 7.2.1-4 1 of 5 shows the Power-Range Neutron Flux Trip Low Setpoint and High Setpoint, and the Intermediate Range High Neutron Flux Trip, and the Source Range High Neutron Flux Trip, all being used on events being initiated f rom a "subcritical" condition. However, Table 3.3-1 shows that'except for the Source Range Neutron Flux items 6b and 6c, all the Trips are inoperable in the subcritical MODES 3 through 5. Further, there is a note d) in the column entitled Tech. Spec (c) of Table 7.2.1-4 which states that "A technical specification is not required [for the Intermediate Range High Neutron Flux Trip and Source Range High Neutron Flux Trip] because the trip function is not assumed to function in Accident Analyses. Please note further that this position is followed Srough in Table 3.3-2 Items 5 and 6 in that a response time is not provided for the Intermediate and Source Range Neutron Flux trips, because it is pro-posed as NA (Not Applicable). Please evaluate the apparent paradox that the Source Range Trip is the only. nuclear Flux trip required to be OPERABLE in the subcritical MODES 3 through 5, and yet there is no Tech Spec proposed for it. At this moment, absence of OPERABILITY requirements for the Power Range Neutron Flux Trip, Low Setpoint, in MODES 3 through 5 would appear _to constitute a disparity with the Licensing Basis FSAR and in a less than conservative manner. 4 The Licensee shall evaluate and propose, those safety-related neutron Flux trips ' which would be appropriate to use and available to trip the reactor for any of those events causing a return to power and under circumstance in which a safety injection initiator is not available, during MODES 3, 4 and 5; and provide the related set Points, Allowable Values and Safety Analysis Limits. Alternately, G the Licensee shall define ana T.S. those conditions and parameters in accordance , with 10 CFR 50.36, which would prevent any such event occurring. , i 06/01/84 18 Revision A

t Please evaluate the conformance with 10 CFR 50 App. A, GDC 20 and 22 of using the Source Range Neutron Flux as a non-diverse reactor trip under cir-cumstances in (MODES 3 through 5) in which there is no Technical Specification i on movable' control assemblies, and which instrumentation consists of only two j channels. Also for circumstances in which all normally available other backup j trip functions such as pressurizer pressure - high and low, and water level I high and " low reector coolant flow", are not specified to be OPERABLE in l Table 3.3-1. The Licensee shall propose on the basis of this evaluation. Items 7 & 8 Overtemperature AT and Overpower AT. The current T.S. provides for operability of these trips in in MODES 1 & 2, and j not 3. ! Occurrences using these reactor trips include events which can be initated from subcritical Zero Power in MODE 2 (Reference 5, Table 7.2.1-4 and Reference 7, Table 15.1.2-2). With the proposed T.S. in which no difference in Reactivity l Condition k f and Shut Down margin is required between MODES 2 & 3, how can ! theLicense8fustifyremovalofthesetripsonentryintoMODE3inwhichthe only difference in RCS conditions is a marginal' reduction in temperature, from the Programmed No Load T aq' Item 11: Pressurizer Water Level - High Operability considerations from MODE 2 down to and including water solid con-ditions in the RHR f10DE are discussed under Section 2.1.118 c(ii.) with a proposal that exclusion of this trip for all these MODES is non-conservative in respect to 10 CFR 50, GDC 20 " Protection System Functions" both for reactivity control systems and overpressure protection systems. The necessity for this trip is increased when reviewed against the totality of the proposed exclusions for Reactor Trip System Instrumentation discussed in the following section under i.tems 2-21 (selected). Items 2-21 (selected): Items 2, 5 and 6: Power Range, Intermediate Range and Source Range Neutron Flux Trips Item 9: Pressurizer Pressure - Low . Item 10: Pressurizer Pressure - High - Item 11: Pressurizer Water Level - High Item 12: Low Reactor Coolant Flow Item 14: Undervoltage Reactor Coolant Pumps Item 15: Underfrequency Reactor Coolant Pumps Item 21: (Proposed) Reactor Coolant Pump Breaker Position Trip. 06/01/84 19 Revision A

e At this time, in MODE 3, 4, and 5, the proposed 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, Low Setpoint; no insertion limits on movable control assemblies, Reactor Coolant ' Pump (RCP) operability requirements permitting less than four (4) RCPs in operation, a Boron Concentration Control which provides less shutdown margin capability than the FSAR requirements, no trip of RCPS on Loss of Flow or Undervoltage or Underfrequency or Opening of RCP breakers, and in addition it is proposed that no trip be provided for Pressurizer Pressure-High,-Pressurizer Pressure - Low, and Pressurizer Water Level - High. And for these circumstances we have no well defined evaluation as to why these reduced projections adequately protect the plant against any of the appropriate Condition'II, III and IV occurrences in these MODES except a Large and Small Break LOCA, and Steam Line Break. We realize the interdependence of many of these. factors in setting a minimum acceptable level of Reactor Trip Protection and that relatively simple solutions are possible, but at this time we do not have available an acceptable analysis and evaluation justifying the proposed T.S. position. The Licensee shall provide an analysis and evaluation of the circumstances under applicable Conditions II, III and IV occurrences in MODES 3 through 5 for an appropriate set of Technical Specification requirements, to ensure conformance to Acceptable Regulatory Criteria and from this he will establish an appropriate range of Reactor Trip System Instrumentation to Safety Related Requirements. The 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 report. Items: 12 Low Reactor Coolant Flow Trip l l 14 Undervoltage - Reactor Coolant Pumps 15 Underfrequency - Reactor Coolant Pumps i 21 (Proposed) Reactor Coolant Pump Breaker Position Trip All these Reactor Trip Functions concern potential for a loss of Reactor Coolant Flow. The proposed T.S. deletes all operability requirements -in MODES 3 through 6. [It also deletes in MODE 2, but this has been discussed earlier under TABLE 2.2-1 items 18.b.a and 12a and 12b]. We have discussed our related concerns and requirements for analyses and evaluations in MODES 3, - 4 and 5 under Items 2-21 (selected) above. A loss of Coolant Flow in the RCS places the plant in an Emergency Operating Mcde. Please advise therefore why such an event should not automatically trip the Reactor in MODES 3 through 5 with the Boron Concentrations being considered for the proposed Technical Specifications. Why should we not use the reactor i trip as a device to ensure complete shutdown of all movable control' rods during any time that a minimum set of RCPs in accordance with operability requirements of the T.S., are not available since RCPs may be required for accident mitiga-tion in MODES 3 through 5 as appropriate. The Licensee shall evaluate and propose. 06/01/84 20 Revision A

Item 13: Steam Generator Water Level - Low Low: Why should not this be required for MODES 3, 4 and 5 (with closed loops) to df embrace the possibility of a return to nuclear power under these conditions. 6tsB) Further, Steam Generator Operability is also required in these Modes to remove decay heat, and Low-Low level alarms are derived from the steam generator low-low instrument channels. Reference 5, Figure 7.2.1-1. The Licensee shall evaluate and propose. Item 17: Safety Injection Input From ESF. See our comments on Table 2.2-1, Item 17 on a proposed revised description for this term to " Reactor Trip From ESFAS. The proposed T.S. proposes that Reactor Trip on ESFAS (or S.I) is not required to be OPERABLE in MODES 3 and 4. Why is reactor trip not required in these dh MODES when Table 3.3-3 for ESFAS Instrumentation, and more particularly Func- Ase) tional Unit 1, including Reactor Trip, shows operability requirements down to and including MODE 4. Further, the licensing basis provides that SI, including reactor trip, be initiated automatically and manually down to MODE 4; see Licensing Basis information in later Section 4.5, EMERGENCY CORE COOLING SYSTEMS, under GENERAL, of this review. . This proposed T.S requirement is therefore non-conservative with respect to the Licensing Basis which requires 'that Reactor Trip on ESFAS (or SI) be Operable in MODES 1, 2, 3 and 4. The Licensee shall evaluate and propose. The Licensee shall evaluate the safety consequences of the fact that in the di event of a Main Stream Line Break'below the P-11 interlock, Reactor Trip will ggggy not be initiated by the Negative Steam Line Pressure Rate - High signal. If the break is outside containment is there is no other parameter remaining which will cause the reactor trip; if the break is inside containment will Containment Pressure-High initiate reactor trip within an acceptable time. What are the consequences of a small to intermediate size break inside containment where, such Contai.7 ment Pressure - High may not occur. We appreciate that Source Range and Intermediate Range Nuclear Flux trips could trip the reactor under these circumstances, on any return to power, but their current proposed status as not being necessary for protection because they are not required in the Safety Anal-yses would leave only the Power Range Low Setpoint Trip, and related resulting power levels of 35% as a Safety Analysis Limit would be unacceptable without a substantive analysis of the event. Please comment in terms of Reactor Trip System Instrumentation Requirements to meet these circumstances. The proposed - T.S is non-conservative in r93pect of Regulatory Requirements in meeting these circumstances; the Licensee shall evaluate and propose. _ Item: Concerning Proscribed Values For % RATED THERMAL POWER DURING STARTUP (MODE 2) AND POWER OPERATION (MODE 1) We note that operability requirements for Reactor Trip System Operation when , expressed in ternis of MODES 1 and 2 are inaccurate and do not represent the ' 06/01/84 21 Revision A

1 1 actual' situation at the plant. T.S. Page 1-9, Table 1.2 defines Power Opera- . tion (MODE 1) at > 5% Rated Thermal Power and Startup (MODE 2) at < 5% Rated - Thermal Pcwer. 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% i Rated Power Level. Further, in this review, under Section entitled TABLE 2.2-1, REACTOR TRIP SYSTEM INSTRUMENTATION SET POINTS, item 18 c(iii) we have identified the need for Safety Analyses Limits for P-10, P-13 and in combination 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% i on the Set Point of 25% to give 35% as the conservative outer limit. If this ! same (total channel error) margin was applicable to both the P-10 and P-13 J channels to give a P-7 Safety Analysis Limit of 10% + 10%, i.e, 20% RATED l THERMAL POWER, then the importance to related safety-related i nues is ( substantively increased. )

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

1) A non conservative discrepancy between the fundamental maximum T.S. Limit of 5% power level in MODE 2 as given on T.S Page 1-9, Table 1-2 and the l nominal value of 10% with a real Safety analysis Limit of.10% plus a Total Channel Error as yet unspecified. 1

2) The elimination of most reactor trip Functions (and many ESFAS Functions) j at this non-conservative power level without a separate comprehensive l Safety Evaluation with respect to Regulatory Requirements and the existing l Licensing Basis.

The Licensee shall evaluate, including our concerns enpressed above, and propose.

                                                                                                                                                      -l l l

06/01/84 22 Revision A

1 j TABLE 3.3-2 REACTOR TRIP INSTRUMENTATION RESPONSE TIMES l

       .            Item 1:         Manual Reactor Trip                                                                    !

At this time, the licensee proposes that the Response Time (RT) for manual reactor trip is not required by safety analysis. Furthermore, he proposes that j in MODES 3 through 5, the only remaining operable trips are those using Source i range neutron Flux and they also are not required by Safety Analyses. Under TABLE 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 'l safe condition in respect to Condition II, III and IV Occurrences in MODES 3 l through 5. Until this evaluation and proposal are accepted, the Licensee ' shall have a Safety Related Manual Trip System to assist in meeting minimum j Regulatory Requirements in 10 CFR 50, APP. A. III. Protection and Reactivity i Control Systems, and the Licensee shall evaluate and propose as a priority issue. At this time, the proposed T.S is non-conservative in respect to Regulatory Requirements for 10 CFR 50, App. A. III. l 4 Items 5 an,d 6: Intermediate Range and Source Range Neutron Flux Trips. d l (RG8) i kSS) As indicated under item Table 3.3-1, items 1-5, these items are proposed as j not being protective actions necessary for the FSAR. Analyses already requested- i will provide a base for determining whether those trips are necessary to pro- 'j tect the plant in MODES 3 througt.1 5. If so, please provide.the necessary techn-ical specifications for these response time in conformance with 10 CFR 30.46. If these values are not provided, all related return to reactivity events shall ) l be evaluated by the Licensee with current FSAR requirements for the Safety Analyses Limit of the power range, neutron flux, low setpoint trip which will , be required to be OPERABLE. The current proposals for these trips is non-conservative with respect to I other proposals in the T.$; the Licensee shall evaluate and propose. l

                                                                                                              "          1 Item 8:    Overpower AT.

l No response time is provided by the Licensee who proposes that a T.S. on this I is Not Applicable. _ Please comment on the fact that this reactor trip is proposed in Reference 5 Table 7.2.1-3 (3 of 5) as applying to five (5) separate Condition II through IV licensing basis occurrences. Also that Reference 5, Page 7.2-14 Rev.42, item 1 d) specifies a maximum of 6.0 seconds (including a transport time of 2 secs) and which is confirmed by Reference 7, Table 15.1.3-1 [alongside Overpower AT]. l The proposed T.S is non-conservative with respect to the Licensing Basis. The l Licensee shall evaluate and propose. Item 9: Pressurizer Pressure - Low 06/01/84 23 Revision A

                                                 +                                                                      .

[

l Item 10: Pressurizer Pressure - High

l i

The TS specifies a Response Time of 12.0 secs. Reference 7, Table 15.1.3-1 ' provides a time delay of 2.0 secs for these events which conflicts with a value of 1.0 secs in Reference 5, page 7.2-14, rev. 42, item 1(e). The Licensee shall clarify. 3

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

[U 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 l function in a Condition IV occurrence in Reference 7 page 15.4-13, item 16 c. I Additional protective functions are discussed earlier under Table 3.3-1, ! item 11. Reference 5, page 7.2-14, Revision 42, Item 1 f provides a reactor trip re- q sponse time at 1 sec. ' Reference our earlier review under Table 2.2-1, item 18.c.(ii). In view of the above information, the proposed T.S. is non-conservative with j respect to the Licensing Basis. The Licensee shall evaluate and propose. {

O Items 8 & 11 General " Although the above two items are not apparently t'he primary reactor trips used as the basis for calculating protection in the Accident Analyses in reference 7, those Analyses represent a limited number of events which are proposed as

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

There is no guarantee that the large number of other possible events will never use these two protection items to primary advantage. Item 16, Turbine Trip - A response time for Reactor Trip on Turbine Trip is not provided in the Technical Specifications. Reference 7, Table 15.1.3-1 advises that the re-sponse time for such a trip is 1.0 sec. but that it is not applicable to the analysis used. _ Reference 7, Section 15.2.10.3, concerning Excessive Heat Removal Due 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 turbine trip and reactor trip are actuated when the steam generator level i reaches the high-high level set point." Also, for the Occurrence of " Inadvertent Operation of the ECCS During Power Operation under reference 7, Section 15.2.14.3, page 15.2-40, revision 43, under Conclusions states that: "If the reactor does not trip immediately, the low pressure reactor trip is actuated. This trips the turbine and prevents excess cooldown thereby expediting recovery from the incident. 06/01/84 24 Revision A

l i 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 10CFR50.36 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.

                                                                                                                             ~
         'The proposed T.S. states that the response time requirement is NA (Not Applic-                                         d b     able). This is incorrect as a separate Reactor Trip is an essential part of                                          gg.p      }

all ESFAs functions during which safety injection is initiated. The required gg) informatio1 is in fact supplied in T.S. Page 3/4 3-30 Table 3.3-5, under the already revised headings proposed above, reference items li, 2b,.3b, 4b. j l This table, un' der response time, should replace the description as recommended l above and alongside each, reference the entry in T.S. Table 3.3-5. ! The response given in the Technical Specifications (except for Manual actuation 6 of SI) are quoted as < 2 secs. No docketed information is available on what values were used in accident analysis, and particularly for MSLB, SBLOCA and (gggj ) LOCA events. The licensee should provide this information and confirm its ! conservatism against the T.S. value, eg. reference 5, Table 7.2.1-4 (5 of 5) ! and related note e. on page entitled " Notes for Table 7.2.1-4" confirms that l Pressurized Low Pressure - Low Level is the first out trip of Safety Injection l for the event of " Accidental Depressurization of -the Main Steam System." The licensee shall explain this terminology - whether we have Reactor Trip on Pres-surizer Pressure - Low which is available at the maximum power output at which !' this particular event is evaluated, or Pressurizer Pressure - Low (Safety Injection) and provide tne associated response time to validate proposed T.S. values. Item 21, Proposed (Reactor Coolant Pump Breaker Position Trip) i As discussed earlier, under table 2.21, Item 14, this trip is provided as an j adjunct to Undervoltage - Reactor Coolant Pump Trip. The Licensee shall l evaluate and propose, f 06/01/84 25 Revision A

l TABLE 3.3-3 ENGINEERED SAFETY FEATURES ACTUATION SYSTEM (ESFAS) INSTRUMENTATION Item 1: Safety Injection, Reactor Trip,.Feedwater Isolation, Component l Cooling Water, Start Diesel Generators, and Nuclear Service Water. . 1 i This description ~of Item 1 lists the various functions initiated by given signals'(which are generally those initiating SI). 3 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

        " Turbine Trip", " Trip of Feedwater Pumps", "Close All Feedwater Isolation                                              {

Valves" and "Close the Feedwater Main and Bypass Modulating Valves. The' term Feedwater Isolation is therefore an inaccurate term to use. It should be removed from this descriptor and replaced by the four separate. functions, as-each of them can be initiated separately and or together dependent upon the ]

       ' initiating Logic.

Further we also note that this functional unit is also that initiated by Steam i 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. l Further, the function to be initiated by Steam Generator Water Level - High j High is function 5 of the.same Table which is again incompletely described and i should be changed (see item 5.later) to. clearly identify these same 4 elements. l Under.these circumstances, the current description for Item 1 should delete l the term "Feedwater Isolation" and Item 5 (see later) should be expanded to l include an additional Functional Unit identified as safety Injection, j l Additionally, the Function " Annulus Ventilation" needs to be added to the descriptor (reference 5, figure 7.2.1-1 (8 of 16) revision 34). i Also, the function unit description " Nuclear Service Water" should include ] [ isolation and startup) of Nuclear Service Water. I 1 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 1c: Containment Pressure - High/ Applicable MODES 1, 2, 3. The Current T.S. does not provide for initiation of SI on 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 MSLB occurs inside containment [so that MSIV Isolation on

       .06/01/84                                  26                               Revision A' t

4 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 MODES 3 and 4. 1his MODE 4 Operability 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 from the existing Licensing Basis, or include the operability requirement in his T.S. The proposed T.S. position is non conservative. Item Id: Pressurizer Pressure-Low This is the same title as used for Reactor Trip on Pressurizer Pressure-Low. This particular/ESFAS actuation is set at a lower pressure and should be described as: Pressurizer 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 Interlock) setpoint. avg I g 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 Pressurizer Pressure - Low (SI) as for Steam Line Pressure - Low (SI) (and implementation of Negative Steam Line Pressure Rate) 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 condition the manual blocks relating to the steam line; if other parameters are used, the Licensee shall evaluate and propose. The Licensee shall also advise of other parameters which may be used to condition the manual block of Pressurizer Pressure - Low (SI). If the Table 7.3.1-3 (1 of 2) and (2 of 2) is correct, then condition MODE.3## should be changed to condition MODE 3# which becomes the correct - description. . Item 2c: Containment Pressure-High-High. Operability is not required in MODE 4. This should be required to be consistent with the evaluation under Item 3.b.3. below. Item 3.b3): Containment Phase B Isolation on Containment Pressure - High High Operability of this isolation is not provided in MODE 4. The Licensee should G advise why this is not necessary for safety when the previous item No.l.e. 06/01/84 27 Revision A i _ _ _ _ _ _ _ _ _ _ _ _ _ _ \

4 1 1 showed reference in the Licensing Basis of protection against Steam Line 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 Operability of Auto-Actuation Logic is required through MODE 4 (Contain- - ment Pressure-High 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 i 7.2.1-1 (8 of 16), revision 34. The Licensing Basis for this requirement lies l inside the analysis of consequences deriving from accidental events whilst the Purge and Exhaust Isolation Valves are open. [Refce CSB] ; The' proposed T.S. is non-conservative with respect to the Licensing Basis; the Licensee shall evaluate and propose. Item 4, Steam Line Isolation 4b: Automatic Actuation Logic and Actuation Relays I The proposed T.S. does not require Operability of Steam Line Isolation Auto Actuation Logic in MODE 4. However, this will be required if the Operability l

                                                 . requirements of Steam Line Isolation on Negative Steam Line Pressure Rate -                                      -

i 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] i

1) System .

2) Individual '

Operability requirements for manual initiation of Steam Line Isolation are not i required by the current T.S. in MODE 4. This however will be necessary to 4 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 4 Operability requirements are given as Y DE 3 and 4. MODE 3 should be con-ditioned as MODE 3# indicating it is only available below P-11 Interlock. l The Licensee shall evaluate and propose. l l l l l 06/01/84- 28 Revision A l

Item 5: Turbine Trip and Feedwater Isolation Reference earlier Item 1 in which this title for Item 5 should be more accurately described as " Turbine Trip, Trip of Feedwater pumps, Close Feedwater Isolation Valves, Close Feedwater Main and Bypass 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]/APPLICA8LE MODES 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 which are expected from the Licensing Basis (reference later Section 3/4.5, Emergency Core Cooling Systems, under GENERAL). The proposed T.S. is non-conservative with respect to the current Licensing Basis and the Licensee shall evaluate and propose. Item 5b: Steam Generator Water Level - High High [to effect Turbine Trip, Feedwater Pump Trip, Closure of Feedwater Isolation Velves and Closure of Feedwater Modulating Valves]/ APPLICABLE MODES 1 & 2. The Licensee should evaluate the need to extend the operability requirements

       -     of this functional unit from current MODES 1 and 2 down to and including MODE 4.

The determining factor may be the availability of Main Feedwater Pumps during these MODES. Plant Operating Procedures which permit Main Feedwater Pumps to be available can cause An Excessive Heat Removal Due To Feedwater System Mal-function and/or Steam Generator overfill unless Safety Related isolation at the Main Feedwater [ containment) isolation valves is incorporated into the T.S. The signal Logic inputs: of reference 5, figure 7.?.1-1, (13 of 16), revision 34, involving Steam Generator Hi-Hi P-14, Safety Injection, Reactor Trip P4, and Low T, would need to be carefuily reviewed, especially since there is currently little or no Safety Related Reactor Trip Protection in MODES 3 ' through 4 so that reactor trip P4 may not be available in conjunction with Low T, g (during cooldown) to effect Feedwater Isolation, and Closure of Modulating Valves, as an inbuilt protection against such circumstances. The proposed T.S. does represent a non-conservative position in respect to the - Licensing Basis, as there is no prerequisite that Main Feedwater is isolated at the Containment Isolation Valves as an LCO, during MODES 3 and 4. The Licensee shall evaluate and propose. Item Sc (Proposed): Safety Injection [to effect Turbine Trip, Feedwater Pump Trip, Closure of Feedwater Isolation Valves and Closure of Feedwater Modulating Valves]/ Applicable MODES 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 1C and Item 5. 06/01/84 29 Revision A l

OPERABILITY is required in all Modes 1, 2, 3, 4, because SI protection has been found neces:ary within the Licensing Basis. The protection was already ' intended in the proposed 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 limited 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 available in the event of failure of one 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 l Actuation Logic and Actuation Relays, and'all related ESF equipment so i 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 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 initiation. Reference also proposed T.S. pages 3/4 4-3 for requirements i for operable RCS systems in MODE 4. The proposed T.S. items 7a through 7g , are generally non-conservative with respect to the Licensing Basis in this ' matter. The Licensee shall evaluate and propose on each of these items, j including consideration of our related review. ; Item 7.a: AFW/ manual initiation Item b: AFW/ Auto Actuation Logic and Actuation Relays: -I 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 4 initiation of Automatic Actuation Logic and Actuation Relays: reference l General above. Item 7.c.1: Start Motor Driven Pumps: Should be operable in both MODES 4 and 5 and especially to counter non-availability of Turbine Driven Pumps early into MODE 4 during the cooldown. 06/01/84 30 Revision A

c l

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

Should be operable in 4. Although nut capable of operating at lower tem-peratures of MODE 4, and MODE 5, it should nevertheless be available for I use to counter consequences described in " General" above, including a station blackout. Item 7.d): Auxiliary Feedwater Suction Pressure Low: - l 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 j unit],on 1 4 Low Auxiliary Feedwater Suction Pressure [This is the parameter caus- ) ing the change] l Operability requirements should identify how many AFW pumps are required to be " tripped" deficient in suction, to effect re-alignment. 1 The licensee should identify those instrument / control channels, and partic-ular engineering alignments, which result in a re-alignment of redundant 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 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 Modes 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 and 5 to meet the FSAR requirements described in " General" above. 1 Item 7.e: Start Motor-Oriven Pumps (by Safety Injection) Applicable Modes have not been identified. NRC proposes MODES 1, 2, 3 and i 4 and 5 to meet the requirements of Item 7: General, discussed earlier, j

Item '7.e: Start Turbine-Oriven Pumps (by SI) - This functional unit proposes that the Turbine Driven AFW pumps are started

      $           by the SI signal. This conflicts with reference 5, Fig. 7.2.1-1 (15 of

16) I&C system Logic Diagram where the initiation of the turbine driven pumps 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 l T.S. has been corrected to show that the Turbine Driven AFW pump does not !

start on Safety Injection.] The Licensee shall clarify. 1 d 06/01/84 31 Revision A i

                                                                                .-.        --_________________ - ____ D

                                                                         "\                          s                                                   :
                                                                                                                                                        \

Item 7.f; Station Blackout - Start Motor Driven 'and Turbine Driven Pumps: . l 1 Provision for. operability is only in applicable MODES 1, 2 and 3. Con-sistent with previous considerations, operability should be required in - j MODE 4, with provision for immediate' operability from MODE 5. j Item 7.g: Trip of Main Feedwater Pumps (MFWP) -_ Starts Motor Driven. Pumps h_ .i 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 all main feedwater pumps is required. The licensee should evaluate' l and propose. l - ] , Applicable modes: The current T.S. proposes Modes 1 and 2#. Condition 2#- l 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 should be 3#. The licensee should explain and l propose. l Please advise why this limitation a't MODE 2 [or 3]# is proposed and how it j 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: j i This is limited in Applicability to MODES 1, 2, 3 by the proposed T.S. i 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 q for, and the consequences of, not proposing this OPERA 3LE IN MODE 4, and I , not being available in MODE 5, to counter'the consequences of potential l LOCAs and loss of RHR cooling in these MODES. The proposed T.S. is ! l non-conservative with respect to the Licensing Basis; the Licensee shall ( evaluate and propose.

l Item 9: Loss of Power: Emergency Bus Undervoltage - Grid Degrade Voltage: Item 9: General. . The Licensing Basis FSAR, reference 7, Section 15.2.9 under LOSS OF 0FFSITE 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 Table 3.3-3. i Absence of this information makes the proposed T.S. non-conservative. 4 The Licensee shall evaluate and propose. ' What does this functional unit do. Please explain, and how many busses to , be tripped for the action to be defined. If it is meant to initiate AFW: l what pumps etc., and if so operability r. 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. 06/01/84 32 Revision A !

m .

                                                                                                                        )
                                                                                                                          \
                   ,         Item 10.a)a.:     Pressurizer Pressure P-11:

Applicable MODES are 1, 2, 3.

Explain the consequences of this non-operability in MODE 4 on availability of dependent protective actions, e.g. , main steam line 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 Licensee shall evaluate and propose. Concerning P-11 Interlock and AFW pumps. The basis provided on proposed T.S. Page B 3/4 3-2 states that:

l "P-11 (i.e. , on. system pressure increasing to P-11 valve) ---- Defeats ; the manual block of the motor drive'n AFW pumps on trip ~of the main feed-water pumps. and Low-Low Steam Generator level." The following information provides the current -Licensing Basis on the. particular proposed interlock P-11 in respect of AFW Pumps: . i The Table 3.3-3, Item 7.c.1, in reference 5, for start of motor driven AFW i 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, Table 7.3.1-3. , Rev-35, ' Interlocks for ESAS - and Figure 7.2.1-1 (15 of 16), revision 34, I&C Logic Diagram. Reference 5, Section 7.4.1.1 6 describes this action under " Bypasses and Interlocks" and that whenever it is' present, an alarm exists in the Control Room. This allows the operator to stop AFW pumps during shutdowns. Supplement No. 5, reference 15, page 22-22 evaluates the use of the P-11 inter-lock as described in the above Basis and concludes that the situation is acceptable. However, the basis for the SER Supp 5 conclusion was that a possi-

ble steam line rupture or feedwater line break were not likely to occur in the proposed MODES 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 not likely to occur has been rejected for these MODES 3 to 5, and detailed atten-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 has been present on other protective actions, the consequences have been fully evaluated. There has never been a related evaluation on the absence of auto-initiation of motor-driven AFWS as now proposed. If the Licensee wishes to pursue this he should evaluate all the events considered in the FSAR below the P-11 setpoint with manual initiation of MD AFW and making due allowance for all the relative reduced and changed projections 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 detailed review of this T.S. has been based on this availability. 06/01/84 33 Revision A

1

                                                                                                               . i
  .                We note that in his submittals concerning this matter, dated March 9, 1981                       l concerning TMI items, the Licensee states that the turbine driven auxiliary feedwater pumps do nct 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 standby'cooldown to achieve. The result is that there. j would be absolutely'no automatic' supply of feedwater for any event beyond ! approx. 20. min into cooldown.

1 At this time, the current Accident Analyses in the Licensing Basis FSAR 1 support the necessity for not using the current bypass for the Motor-Driven ' Pumps.

                                                                                                                   -j l                   The Licensee shall advise what safety-related reasons require that he must

bypass automatic startup of the motor-driven auxiliary feedwater pumps on l top of both main feed pumps, and on SG Low Low-Level in the final stages of: l. plant shutdown. Also, what prevents him from installing automatic restoration on receipt of the related protection' signal.

1 Item 10.b; Interlock; Low-Low Tavg P-12: 4 Applicable MODES are 1, 2, 3. ( Reference Item Table 3.3-4, Item 10b, of this document. Since Interlock P-12 effectively provi, des and limits steam dump capability, including accidental blowdown, by constraining it to 3 cool down dumps.to i the condenser; why remove this interlock in MODE 4 and MODE 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 Licensee shall evaluate and propose. 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 absence of this will affect the various functions to be performed and how they will impact the FSAR requirements for plant protection in MODES 4 and 5. This should be for both the " Reactor tripped" and " Reactor not tripped" condi-tions considering that the reactor can be in both situations during these d Modes. Licensees evaluation to items Sa, b and c above should.be also considered in this evaluation. - 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 evaluation and p'ropose. Until Safety Related Isolation of Main Feedwater l 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. Jtem 11 proposed: (g, There is a need to add a new Functional Unit not addressed in the current T.S., but which is a part of ESFAS. Thi's 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 16) revision 34 for the related l l unique control logic. This Function is initiated by: lla. Reactor Trip P-4, and Low Tavg. lib. Reactor Trip P-4, and Steam Generator Level - High High P-14. 11c. Steam Generator Level - High High P-14 (see 5 above) lid. Safety Injection (See 5 above). Operability for lla would be in accordance with 10c (above) and later evaluation under Table 3.3-4 Item lla (Proposed). Operability for 11b would be in accordance with the evaluations in 10c and d above, l Operability for 11c and 11d would be by reference to items 5, Sabc. 1 TABLE 3.3-3: TABLE NOTATION. I: 6r The uncertainty of the notation under ## is discussed in Item le earlier. Please amend as required in accordance with the related resolution. 1 l l l l l l l 06/01/84 35 Revision A

e l TABLE 3.3-4: ENGINEERED SAFETY FEATURES ACTUATION SYSTEM (ESFAS) _ INSTRUMENTATION TRIP SET POINTS General: These have been checked against the information in reference 18, table 3-4 and related NOTES FOR TABLE 3-4 on page 3-13 and which is de- i scribed as being applicable to McGuire Unit 1, 50-369. At this time, the l assumption is made that this information also applies to 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. 3 a, I ItemNo.Ig: ! 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 1c: 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 wh'y 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 shall evaluate and propose. Item 4d: Negative Steam Line Pressure Rate - High [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 to be

         "-110 psi"; an additional descriptor is also necessary reading: "with 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 conservatism, 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 psi and 110 psi. This is also internally consistent with the descriptor in Table , 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 !

l i

J Please discuss the logic of the values ~in reference:18. A Trip Set Point 3 of a negative rate of 110 psi with an allowable value.of 100 psi..(both. I with a time constant of 50 psi)'would provide that an earlier isolationf of the MSIVs is less, conservative, and this is.not so for the MSLB event. The expectations are that negative rate for the allowable value would be higher than for the Set Point. Please clarify- " i Further, the same reference 18 Table 3-4, column'12, states under 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); i please describe how the plant is otherwise protected through the proposed T. S. Otherwise, please provide analyses which show that the plant is pro- - tected by this proposed setting under proposed T.'3. requirements. This-item is related to our other concerns on Technical Specifications on-Bora-tion Control under earlier Section 3/4.1.1 Boration Control. The proposi- ; tion that this value is not used in Safety Aanlysis is non-conservative. ! The Licensee shall evaluate and propose. Item 5: The description of this Functional Unit should be revised and j 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 ; Table 3.3-3, Item Sc) ! Item 6a & b. Containment Pressure Control System The licensee should provide the basis for these Set Points and Allowable Values. - l t Item 7(c): Steam Generator Water Level - Low-Low The licensee should respond to.our concern under Table 2.2-1, item

13. -.

Item 7(d): Auxiliary Feedwater Suction Pressure Low The description should be revised as proposed under our earlier Table 3.3-3 item 7d. Provide the basis for the values given. Items 7c(1) and (2): Concerning start of Motor Driven and Turbine Driven Pumps This technical specification provides that the motor-driven AFW Pumps start on low-low in one SG whereas the turbine driven pumps require low-low in-two SGs. This appears to be in conflict with the accident evaluation 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

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

Reference 5, Sectinn 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 gpm to three intact steam generators while feeding one faulted generator and both motor driven pumps together can supply 450 gpm to three intact steam generators while feeding one faulted j 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),             ,

states: "The single active failure assumed in the analysis is the l turbine driven auxiliary feedwater pump. The motor driven pump that ! is headered 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 gpm to each. This yields a total flow of 450 gpm to the intact steam j l generators one minute after reactor trip. At 30 minutes following l the rupture, the operator is assumed to isolate the auxiliary feedline ] to the ruptured steam generator which results in an increase in 1 l injected flow of 80 gpm." l The sequence of events in the accident evaluation in Reference (7), Table 15.4-1 shows that after the accident is initiated at a programmed value of SG level, the low-low SG 1evel in the ruptured SG is reached . 20 secs later, and auxiliary feedwater [at 450 gpm] is delivered to the 1 intact steam generators in 61 sec. i l It appears, based on the above information, that on SG low-low in the ruptured SG, both the motor driven and the turbine driven pumps are initiated (with the single failure being in the turbine driven pumps). This is not in accord with the T.S. If it is assumed that low-low level in the other SGs is also reached at the same time by bubble collapse, please justify. We note that the Reactor & Turbine Control System is designed so that under normal operation, collapse of SG level on Turbine - 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. The proposed T.S.s 7c0 and 7.c(2) appear to be non-conservative in respect of Accident Analysis used in the Licensing Bases. The licensee shall clarify, evaluate and propose; this should be in conjunction with our other concerns on this event noted later in Sections of this review. 06/01/84 38 Revision A

n a q :o )

 .                                                                                                                                  t
        -Item 8:    Automatic Switchover to Recirculation'         t 3

The Licensee shall provide the basis for the se',t point values of the RWST j levels specified. What are the allowable values for [ drift and] total ! channel errors and the related Safety Analysis Limit. f Item 9: Loss of Power Confirm the bases for the set points and allowable values specified. . Item: General The Licensing Basis FSAR, reference 7, Section'15.2.9 under LOSS OF i 0FFSITE POWER TO THE STATION AUXILIARIES describes'a set of Reactor 1 Protection System and Engineered Safeguards Features Actuation Responses i for the Plant, to ensure its safety. Why is this particular-set of ESFA's Functional Units and related Instrumentation Set Points not provided in this item under Table 3.3-4? , s 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. l The FSAR has analyzed the consequences of this reduced level of protection for a limited number of these occurrences and this has been based on a system pressure of 1900 psig; Reference 8, page Q212-47, item 212-75 1A. Why then is a trip set point of 11955 psig used. This set point value should be below 1900 psig with appropriate allowances for drift and channel errors to the limiting value used in the Safety Analysis'of 1900 psig. The current specification is non-conservative with respect to the Licensing Basis FSAR & therefore not in accordance with 10 CFR 50.36. The licensee shall provide a safety evaluation for the difference, for approval, or restore the set point to be a valid T.S. value.

Item 10b: ESFAS Interlock T ag -Pa. 4 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 which is' removed and then with the exception of 3 cooldown dump valves (to the condenser). The steam generator Power Operated [ atmospheric] Relief Valves (SG PORVs), are not affected: Please correct the Basis.

06/01/84 39 Revision A

m . A set point of 553-551 F is provided. 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 complete 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, Trip, P-4, with NA (Not Applicable) trip setpoint & Allowable values." However, should not this item read as: 10c. P-4-with Trip Setpoint and Allowable values defined as ~in Reactor l Trip to Table 2.2-1, with the exception of: " Power Range, Neutron Flux, High 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 why R= actor 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 ev.aluate 1 d and propose Item 11 Proposed: There is a nee'd to add a new Functional Unit not addressed in the current' T.S., but which is a part of ESFAS. This is:

                                                                                 "Close Feedwater' Isolation Valves & Close Feedwater Main & Bypass                I Modulating Valves." (See Reference 5, Figure 7.2.1-1 (13 of 16)                   l Revision 34.)                                                                     i l

l This Functional Unit is initiated by;

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

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

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

j d. Safety Injection (see 5 above). " Trip Set Points would be in accordance with the related values in earlier f l Items 10 and 5 of this section. l 06/01/84 40 Revision A

-v Reference Item 11b above, involving Reactor Trip P-4 & Steam Generator High High Level P-14.

The NRC has observed potential situations of concern involving this j 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 feedwater 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 power which will cause a reactor trip from the neutron flux low power, 25%, setpoint,.and 35% (Limiting Safety Value in Analysis) and hence generate )

                'a P-4 signal, but will not correct the initiating cause of the faulted              '

main feedwater control system until SG-Hi Hi level is subsequently ini- ' tiated and effects closure of MFW isolation valves. Whereas the FSAR evaluates the first even,t 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 DNBRs 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 which evaluates the worst case.                        .

NRC Safety Concern C: The licensee.has provided no information on " Safety ! Analysis Limits" that would be applicable to Permissive P-8 in evaluating the above events. If the allowance is ultimately of the same order as for i the Power Range, Neutron Flux - High and Low Set Point Trips, i.e., approx. i

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

1 In respect of NRC Safety Concerns A, B, and C above, we consider tha pro-posed T.S. in respect of the related permissives and in+erlocks to be non-conservative with respect to Regulatory 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 I appropriate. This could be considered a Generic Issue. ! General: In view of the consequences of the bypass of reactor trip on turbine trip below P-8 for the events protected by trip of turbine on 06/01/84 41 Revision A 1 l . - - - - - - - I

y 3

                                                                                                ,     'I Steam Generator Hi:Hi., the licensee should review the analyses for all other Condition 11 through IV occurrences to determine whether the con-              -

clusions~ deriving from the existing evaluations need-to be altered. This . could be considered a Generic Issue. l

                                                                                                        .1 Reference Item 11(a) above, involving-Reactor Trip P-4 and Low T ava, i

Reactor Trip P-4 together with Low-T avg 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 , 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 l Reference 5, Section 7.7.1.7 under the title Steam Generator Water. Level Control, in the following terms:

                   " Continued delivery of feedwater to the steam generators is required 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 tempera-ture, and the reactor has tripped. Manual override of the feedwater              )

control system is available at all times."

      .                                                                                                 1 i             This P-4/ Low T gg combination does perform a safety function.in preventing excessive cooldown after the reactor is tripped, but has never been                        l incorporated, or discussed in the Section 15 FSAR analyses (Reference 7) for this purpose.                                                                          1 Within the FSAR under Reference 7, Section 15.'2.10.1 " Excessive HEAT                      l REMOVAL DUE TO FEEDWATER SYSTEM MALFUNCTIONS" state that:
                   "An accidental full opening of one feedwater control valve with the        ,         i reactor at zero power and the above mentioned assumptions, the maximum reactivity insertion rate is l'ess than the maximum reactivity insertion rate analyzed in Subsection 15.2.1, Uncontrolled Control RCCA Bank Withdrawal from a Subcritical Condition,.and therefore, the results of the analyses are not presented. It should be noted that if the incident occurs with the unit just critical at no load, the reactor may be tripped by the power range high neutron flux trip (low            _

setting) set at approximately 25 percent."

                   "For all excessive feedwater cases continuous addition of cold feed-                 '

water is prevented by closure of all feedwater control valves,'a trip of the feedwater pumps, and closure of the feedwater pump discharge valves on steam generator high-level." This event from zero and higher power levels (already discussed under

   .         earlier Item lib) is initially protected by the high neutron fluxtrip;                     '

however whilst this provides immediata protection, the main feedwater is L not isolated and continue to cooldown the reactor with continued reactivity addition. The licensee must confirm that acceptance criteria for the ; reactor system are not exceeded if further protection must wait for Steam 06/01/84 42 Revision A

7 .

                                                                                                                          . a 1

l 1 I Generator Hi Hi Level "to trip the MFW pumps, and together with existing i

     .         Reactor Trip to provide Main feedwater Isolation. Or, is it necessary to                                        i 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 avg

Inclusion of the P-4 and Low T, g interlock into the T.S. would provide more reliability in protection .for this event in conformance with the diversity criteria of 10 CFR 50 Appendix A, GDC Criterion 22 in support GDC 20. . Without this, there is no diversity for protection from this l continuing event. The proposed T.S. should require T Low to be incor-3yg porated into the T.S. to meet the above Regulatory Criteria. The licensee shall evaluate and propose.

The licensee shall evaluate this issue with our concerns expressed under Table 3.3-4, Item 11 proposed, Reference Item 11(b) above, NRC Safety Concerns B and C to which this is directly related. The presence of Low Tavg, without T.S. considerations of Set Point, 1 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 combine with a ) Reactor Trip from full power to close Main Feedwater. [ containment] Isola- ) tion valves and Main Feedwater Mooulating valves and cause.a more severe l transient than would otherwise be necessary. The Licensee should evaluate * ' the cons' e quences of single f ailure on appropriate Conditions II, III, and IV Occurences, and propose as necessary. Item: Reference 7, Section 15.2.14, page 15.2-38, Revision 43, which is the ) Accident Analysis for " Inadvertent Operation of ECCS During Power Operation," I states that: l l l Spurious ECCS operation at power could be caused by operator error or , l a false electrical actuating signal. Spurious actuation may be assumed l to be caused by any of the following: '

1. High Containment pressure

2. Low pressurizer pressure

 -             3. High steam line differential pressure I

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

l Please explain the signals 3 and 4 since they do not appear in the TABLE 3.3-4 just reviewed, nor do they seem to appear in the Logic Diagrams of the Licensing Basis in the FSAR to reference 5. The Licensee shall evaluate and propose. i 06/01/84 43 Revision A i

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. j i

1 4 l l l 1 l l l

I I e 1 d l l 1 f i 1 1 1 J l l l

! 06/01/84 44 Revision A

7 TABLE 3.3-5 ENGINEERED SAFETY FEATURES RESPONSE TIMES Item 2a: Initiation of Safety Injection by: Containment Pressure-High. A value.of 5 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 I sec. -

                                                   'l 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 upon for diverse protection. However, it is the only automatic initiation of Safety Injection protection below [P-11].- Other events dependent upon a SI generating signal, particularly circumstances descibed-under items 3a and 4a below, shows safety analyses limits of 5 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 licensee shall evaluate this difference and propose accordingly. Item 2b: Initiation of " Reactor Trip (From SI)" by Containment Pressure-High ! The descriptor (From SI), should be deleted as it is incorrect. The response time is give is 5 2 secs and this different from the FSAR, Reference 5, page 7.3-8 which gives a maximum ti'me of.1 sec. This value is less conservative than the FSAR and the licensee shali evaluate and propose accordingly. i Item 2c: "Feedwater Isolation" from Containment Pressure-High i The response time is given as 1 9 secs. Reference 5, page 7.3-8 shows that initiation of ESFAS from this source is i a maximum of I sec. Table 3.6.2 of the T.S. provides isolation times of 1 5 secs for main feedwater containment isolation and < 10 secs for main feedwater to Auxiliary Feedwater Isolation. A total time to isolation of MFW, from l . Containment Pressure-High, of 111 secs seems appropriate to available i equipment, There would then be a conflict between the response time of 1 9 secs in t l the proposed T.S. and the potential value of up to 11 sec from other l licensing basis information. No event in Reference 7, Section 15.1 through 4. uses this particular isolation in time Analyses. However, this is a important factor for containment integrity during a Main Steam Line Break in containment. The value used as the Safety Analysis Limit shall be provided by the licensee, 06/01/84 45 Revision A

e

    .                                                    compared with proposed T.S. Item 2c and any differences evaluated, and                    ,

T.S! proposed as appropriate.

1 Item 2d: Containment Isolation - Phase A, from Containment Pressure-High The proposed T.S. values are 18(3) (with offsite power) and 28(4) without offsite power. I l

Reference 5, page 7.3-8 shows that initiation of ESFAS from this sou'rce l is 1 sec. l Table 3.6-2 shows Maximum Isolation Times of up to 15 secs for Reactor q Coolant Pressure Boundary Isolation valves. A minimum total time to J containment and isolation [for the RCPB] of 16 secs seems feasible, plus l 10 secs giving 26 secs total without offsite power. j i The proposed T.S. values should be checked against those used as Safety I Analysis limits for related Conditions II, III, and IV occurrences using j SI. Values used by licensee shall be provided, comcared with Item 2d. and any differences evaluated. Item 2e: Containment Purge and Exhaust Isolation, from Containment Prc1sure-High This is given as N.A. This is not so; response times have be used to minimize offsite consequences of-any Condition occur' ring whilst contain-f.aht purge & exhaust is being used, ibis 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-High. 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 the licensing basis; the licensee shall evaluate and propose. Item 29: Initiation of Nuclear Serv lce Water (NSW) from Containment Pressure-High This response time is given as 1 65(3)/76(4) secs. The superscript 3 does not seem appropriate; whilst the related Notation on T.S. Page 3/4 3-33 refers to absence of diesel delay (i.e., no loss of offsite power), it describes start up of ECCS equipment'but does not include the requirement for " Isolation-and Startup of Nuclear Service-Water Pumps as described in Functional Unit 1 of Tables 3.3-3 and.3.3-4. The same cornment 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

y _ _ . - 4 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 supplies. In other sec-tions of this review, the licensee has been asked to re evaluate Safety Analyses to recognize this fact. Parallel with this, the licensee shall 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-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 given as 65(3)(3)/76(4)(2) secs. I The description ~ of superscript 2 under Table Notation on T.S. Page 3/4 3-33 is incomplete. The licensee shall propose an accurate description of these 4 circumstances including its dependence on Nuclear Service Water; the ' licensee should confirm that this cooling water supply information is for this safety related service. l . Reference 5, page 73-8 shows the initiation of ESFAS from this source is I sec. No other information is available on' Safety Analysis Limits used in the FSAR. The licensee shall provide this information for related Condi-tions II, III, and IV Occurrences for both on-site and offsite power. This ] information shall e eveluated and the licensee shall propose; At this i time, considering the non-conservative circumstance with NSW AFW supply, it must be presumed that any Occurrence required to utilize the Nuclear Service Water must be considered non-conservative with respect to the values curr'ently presented in the FSAR, Reference 7, Section 15. Item 2i: " Start Diesel Generators" from Containment Pressure-High A response time of i 11 secs is given. Reference 5, page 7.3-8 shows that initiation of ESFAS from the source is a maximum of 1 sec. -, No evaluation in Reference 7, uses this sensor as the prime initiator above the P-11 Interlock, although it is relied upon for protection above, and directly for protection below [P-11]. Other events dependent upon a SI generating signal particularly, items 3a & 4a below, show safety analysis limits of 5 10 secs for this value. In respect of current safety analyses limits, therefore, it appears that the proposed value is less conservative than the Safety Analysis Limits. The licensee shall evaluate and propose. 06/01/84 47 Revision A L________ _ .__ ____ o

y 4 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: Pressurizer-Pressure-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 sensor is'< 1 sec (Reference 5, page.7.3-8). This is also~the same initiation time for Containment Pressuro-High. Since both or either_of these initiators can

          'be available in Occurrences involving 51, and initiation times are the same, our comments and conclusions under earlier Item 2 can be direct'ly referenced for items under Item 3 in cases where the proposed response' time is the same fcr a given ESFAS function.

Item 3(a): " Safety Injection (ECCS)" on Pressurizer Pressure-Low [SI] Values of ,< 27(1)/12(3) secs are proposed. Reference 5, page 7.3-8, shows a maximum initiat'ing't4me of ESFAS 1.0 secs for this signal. The value of 12 secs (with offsite power) is consistent with safety analysis limits given for the MSLB in.refa_rence 7, page 15.4-10, Section 7 where "In 12 seconds, the valves are assurtaa 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 star. 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 within the licensing basis of 12 secs. The proposed.value of 27 secs (with LOOP) is however larger than the value of 22 seconds from the reference described above (i.e., 12 secs + 10 secs l delay for start of diesel). This value of 27 secs therefore appears less q l conservative than the FSAR, reference 7, page 15.4-10, and the licensee ;

 .           shall evaluate and propose.                                                        _

Item 3b: " Reactor Trip (from SI)" on Pressurizer Pressure Low [SI) i The descriptor (from SI) is' incorrect and should be deleted. A value of 1 2 secs is proposed. The FSAR in Reference 5, page 7.3-8 quotes a value of i 1 secs. j l The proposed T.S. value appears less conservative than the Safety Analysis j Limit and the licensee should evaluate and propose. l 4 06/01/84 48 Revision A , 1 l

1 Item 3c: " Feedwater Isolation" From Pressurizer Pressure-Low (SI) The proposed T.S. is 5, 9 secs. !

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

Item 3d: " Containment Isolation - Phase A" from Pressurizer Pressure-Low (SI) The proposed T.S. is j,18(3)/28(4) secs. Reference our concents and requirements under 2.d. above. Item 3e: " Containment Purge & Exhaust Isolation" From Pressurizer Pressure-Low (SI) i i i The proposed T.S. is NA. i Reference our comments and requirements under 2.e. above. Item 3f: " Auxiliary Feedwater" Initiation by Pressurizer. Pressure-Low (SI)- The licens'e proposes NA (not applicable). 1- Safety injection logic closes the main feedwater 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 and a related response time. It is outside the licensing basis, not to a propose a value for this response time. This T.S. value is therefore non-conservative; the - licensee shall evaluate and propose. Item 3g: " Nuclear Service Water System" Initiation from Pressurizer Pressure-Low SI The T.S. value is given as 76(2)/65(3) secs. Our comments on 65(3) are as for our earlier 2g. With resoect to superscript (1) on 76; why is this different to Containment

  .                   Pressure High which is 76(3) when the concomitant SI signal gener=tes the              '-

same equipment requirements. Superscript (1) now provides for Si ona RHR pumps whereas (3) did not. Also, superscript (1) , if it is to be used should include Isolation and Start 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 ,

2) (3) and (4) and ensure that they are complete, accurate and consistent with all the related ESFAS initiating signals and functions.

06/01/84 49 Revision A

i i This position appears inaccurate & confusing to the extent ~that it must , be-considered non-conservative. Item 3h: Initiation of Component Cooling Water from Pressurizer , Pressure-Low (SI) The proposed T.S. is 576(1)/65(2)(3) secs. ' See our comments and requirements under 2h. and 3. General above. Item 31: Start Diesel Generators from Pressurizer Pressure-Low (SI) The T.S. value is 1 11 secs. See our comments under 21. above which are substantively applicable to this item. Therefore, the proposed 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 sec 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 licensee shall evaluate and propose. Item 4a: " Safety Injection (ECCS)" Initiation on Steam Line Presaure-Low These values of 5 12(3)/22(4) agree with the Safety Analysis Limits of the Licensing Lasis FSAR. Item 4b: " Reactor Trip (From SI)" from Steam Line Pressure-Low. ThedescriptJop(fromSI)isincorrectandshouldbedeleted. This value of 5 2 secs agrees with Reference 5, page 7.3-8. 1 Item 4c. "Feedwater Isolation" from Steam Line Pressure-Low I 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 tiie potential value of up to 11 + 1 = 12 seconds' from Licensing Basis Information. l

        -06/01/84                                  50                                  Revision A i

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Item 4d: " Containment Isolation - Phase A" on . Steam Line Pressure-Low j The proposed T.S. is 5 18(3)/28(4) secs.

         -                                                Reference our comments and requirements under 2d. above, modified in that proposed I.S. times appear feasible with the' additional delay of I sec.

Item 4e: " Containment Purge and Exhaust Isolation" on Steam Line Pressure-Low l The proposed T.S. is NA. 1 Reference our comments and requirements under item 2d. above. Item 4f: " Auxiliary Feedwater Pumps" initiated by Steam Line Pressure-Low The proposed T.S. is NA. I Reference our comments and requirements under 3f. above. ' tItem 4g: " Nuclear Service Water" initiated.on Steam Line Pressure-Low ; The proposed T.S. is 5 65(3)/76(4) secs. i Reference our comments, requirements, and remarks under 2g., 3g., and 3 l General above. Item 4h: Steam Line Isolation.on Steam Line Pressure-Low. i The propos,ed TS value is 1 9 secs. Reference 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 li m stop valves

                                                   .      are " designed so close in 5 secs...". A minimum closur' af'7 secs seems likely.

For actual safety analysis limits, Reference 7, Table 15.4-1 (1 of 4) and 15.4-1 (2 of 4) both show a difference of seven (7) secs between arriving ] at the " Low Steam Line Pressure Setpoint" and "All main Steamline Isolation Valves Closed." [In the case of Feedwater System Pipe Rupture)

       -                                                                                                                                              -i The proposed TS value of f 9 secs is therefore greater than the Safety                          l Analysis Limit.

The proposed TS must therefore be considered less conservative for this event. The licensee shall evaluate and propose. 1 Item 41: " Component Cooling Water" Initiation by Steam Line Pressure-Low l Proposed T.S. value is 65(2)(a)/76(2)(4) Reference our earlier comments and requirements under 2h and 3h. above. 06/01/84 51 Revision A L_______._.______._.____._ - - - - - - - - - .

e 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 2i'above. . Item Sa: " Containment Spray" - Initiated on Containment Pressure-High-High Licensee shall provide the Safety Analysis Limit and compare with .the proposed'value of 1 45 secs. Evaluate and propose 1as necessary. Item Sb; Containment Isolation - Phase B on Containment Pressure-High-High-This is proposed as Not-Applicable. The licensee should propose why this is so when it appears that TS Table 3.6-2 Containment Isolation valses, Maximum Isolation Time (secs), applies only to closure from receipt of signal, and may not include the ESFAS Response Time. Reference especially T.S. page 3/4 6-30 where main steam line isolation is specified at 5 secs compared with the same value quoted on Reference 7, page 15.4-6 which states that these fast acting steam line valves are designed to close in 5 secs and Safety Analysis Limits have been shown as 7 secs under Item 4h. above. What is needed to supplement the information in T.S. Table 3.6-2 is the ESFAS 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 pressure (related page 7.3-7), and also as 1 sec for closing main steam line stop valves on Containment Pressure-High [High], It appears this

         , item should read as:                                                                    1 Sb. ESFAS Input to Containment Isolation - Phase B            1 sec The licensee shall clarify, identify the related-Safety Analysis Limits, and evaluate as appropriate. Until then, the proposed T.S. must be considered non-conservative with respect to the Licensing Basis, Item Sc:   Steam Line Isolation on Containment Pressure High-High l

The proposed T.S. value is 1 9 secs. Reference 5, page 3.7-8 shows containment pressure initiating ESFAS signals

 ,         with a 1 1 response time. Item 4h. above shows fast acting stop valves                _

closing in 5 secs. giving a total time of 1 6 secs. Since MSIV actuation under Containment-Hi Hi can be caused by MSLB which provides for a maximum of 7 secs above, the proposed value of 9 secs appears less conservative. i A comparison also with values used in assessing environmental releases from containment should also be made. 06/01/84 52 Revision A

                                                        .The licensee shall identify the Safety Analysis Limits used for this Steam
       .                                                 Line Isolation, including the MSLB in containment, evaluate against the           l proposed T.S. value and propose as appropriate. Until such time, the              I current value appears non-conservative.

Item 6a: Turbine Trip on Steam Generator Water Level-High High The proposed T.S. is NA, i.e., not applicable. l l Reference the licensee to our comments under Table 3.3-2, Item 16 where j it is shown that it is used within the Licensing Basis. , i 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. Item 6b: "Feedwater Isolation" Initiated by Steam Generator Water Level-High High - The proposed T.S. is 1 13 secs. j l' 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 l trip" is based on an ESFAS time delay of 2.0 seconds. l 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 depending on this response time is that for the Excessive Cooldown occurrence under Reference 7, page 15.2-28, and for this, no value is quoted for isolation of main feedwater which is the initiator of the event. However, Figure 15.2.10-2 shows,that with ini-tiation of the event caused by one faulty control valve, it takes 32 secs l to reach the SG-High-High Level with a m. ass increase of 35% of initial, and thereafter does not increase further. This implies zero closure time. Since it is 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 Ma.in Feedwater level can affect the consequences of the event at power, if there has been a trip, with a potential for power restoration and/or overfill of the S-G to cause water ingress into the main steam lines. Additionally, it can have consequences of potentially j larger importance for the event occurring from zero subcritical power. Reference also our concerns under item Table 3.3-4, item 11b and lla above. l - The licensee shall evaluate the related concerns, including the extended MFW valve isolation times, to determine their safety significance, and 06/01/84 53 Revision A

propose as required. Until that time, it must be concluded that since a zero (0) value has been used in the current analysis, that the licensee l has a potentially non-conservative situation with respect to Regulatory Requirements of Reactivity Control and Regulatory Concerns for Flooding of the Main Steam Lines. - Item 7a: " Motor-Driven 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 5 60 secs. The FSAR Safety Analysis Limit is 61 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 feedwater delivered to steam generators is 61 secs. The current proposed T.S. value is therefore conservative with respect to the current safety analysis limit. However, the current safety analysis limit of 61 secs currently used appears to be a mistake and not in accordance with Regulatory requirements. 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 8 motor driven pump is aligned to the 8 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 of 13 secs]. Earlier information under this T.S. Table 3.3-5 shows that the response time for Nuclear Service Water Supply is 65 secs, assuming offsite power available and 76 secs assumi.ng 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 occurrences involving AFW supply would need to be re evaluated to establish acceptability. 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, 40 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

m - 1 Item 11, below. If this is confirmed at from 65 to 70 secs,. or any longer I time'.than used as the existing Safety Analysis Limit in the FSAR, then j acceptable re-evaluation of all Conditions II, III, and IV occurrences i involving AFW supply, are required by 10 CFR 50.36. l

       -                                                                                                                                                                      .j Our current evaluation is that the response, times in the proposed T.S.                                                                        I 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 1 9 sec. Reference 5, page 7.3-8 states that the maximum allowable time for l generating the ESFAS MSIV isolation signal from.a Steam Line Pressure ! Rate circumstance is 2 secs, the same as for item 4h. above. 1 l Our comments and requirements therefore are the same as under item 4h. We appreciate that this signal is generated at below P-11, but with.the existing proposed Boration Control T.S. we must continue to evaluate this value as non-conservative. q l The proposed T.S. value is greater than the Safety Analysis Limit uf seven (7) secs and must be considered less conservative for this event. The licensee must evaluate this difference and propose, i

Item 11: " Automatic Re-alignment of AFW Supply on Low Suction Line Pressure" [The existing description should be changed to more accurately state this action) 1 Proposed T.S. value is 13 secs. i I Note our comments under 7a. and 7b. above. Altho' ugh this response time may ) I be in accordance with current plant engineering, it is not in-accordance I

with the existing Safety Analysis Limit for Auxiliary Feedwater Supply l which, on current information, has pre supposed no such transfer time.

If a tank has been lost because of seismic action, we cannot assume a residual 15 secs supply at this time. l 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. Item 12: " Automatic Switchover to Recirculation" on Low RWST Level Response time proposed as 1 60 secs The licensee shall provide the bases for this value and eyaluate against this 1 60 secs, and propose as necessary, , 06/01/84 55 Revision A t l u______ _ _ _ _ _ _ _ _ __ _ _ _ _ _ _ _ _ _ _ _ - . _ _ _ _ _ . _ _ _ _ __ _ _ _ _ _ _ _ _ _ _ _ _ _ _

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-0FF-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 l Response times required to protect the facility on Station Blackout ~condi-

             'tions makes.the proposed T.S. non-conservative with respect to the 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 Blac'out k 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. g Item 14: " Start Motor-Oriven Auxiliary Feedwater Pumps" on Trip of Main 1 Feedwater Pumps Proposed T.S. value is < 60 secs. Reference our comments under 7a. and 7b. above together with the necessity 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 1 11 secs. Reference our comments under T.S. Table 3.3-3 Item 9 and Table 3.3-4 Item 9 and provide appropriate clarification. No evaluation is possible at this time, t . 06/01/84 56 Revision A

v Item 15: Loss of Power Item 15: General

       '            Our review comments under item 13 " Station Blackout" are fully applicable to this item with the related conclusion that:

The absence of most of the information on Functional Units and related Response Times required to Protect the Facility on Loss of Power makes ' the proposed T.S. non-conservative with respect to the Licensing Basis. ) The Licensee shall evaluate and propose. -l Item [ Foot] Note: Response time for Motor-Driven Auxiliary Feedwater Pump Starts on All SI signals. This is proposed as < 60 secs. l Reference our earlier comments for its inclusion in Items 2f. , 31. , and i 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, these values are non-conservative with respect'to Regulatory requirements and the licensee must evaluate and prcpose. i Item: Table 3.3-5, TABLE. NOTATION'on.T.S. Page 3/4 3-33 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 items9 2 ., 2h., 3g., 3h , 4g., and 41. above. Notation 2 of this Table states that: (2) Valves 1KC3058 and 1KC3158 for Unit 1 and Valves 2KC3058 and 2KC3158 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 < 30(3) 3.b < 30(3)/40(#)

4.b ]30(3)/40(#) { Since the functions 2b, 3b and 4b are all Reactor Trip functions, please explain. - Since these descriptors are apparently incorrect, provide the correct descriptors. 1 1 06/01/84 57 Revision A I l

                                                                                                      ,j 1

i Since-superscripts (3) and (4)-used above'make no mention of Component

          . 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                        i Actual overall' response time, and has this been incorporated into~the

i Safety Analysis of the Licensing Basis. I The licensee shall clarify, evaluate and propose. Lack.of accurate .

I information on response times must be considered'as non-conservative, o j l t i l 1 1

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     -06/01/84                                 58                               Revision A

   .                                                                                                          l Section 3/4.4 REACTOR COOLANT SYSTEM Section 3/4.4.1 REACTOR'C00LANT LOOPS AND COOLANT CIRCULATION l

Item: GENERAL 1 G.1 ' INTRODUCTION , j a Concerning RCS' Operability requirements, in M'0DE 3-5: i We refer to our earlier discussions & licensee requirements - and especially ! under Section 3/4.1.1, T.S. Page 3/4 1-1, 2 & 2a on Boration Control, T.S., j Page 3/4 1-20 & 1-21 concerning SHUTDOWN AND CONTROL ROD INSERTION LIMITS and j TABLE 3.3-1 REACTOR TRIP SYSTEM INSTRUMENTATION generally, including ~more- i

          'particularly items 2-21 (selected) and items 12, 14, 15 and 21.                                j 4 Under our item T.S. TABLE 3.3-1, items 2, 5 & 6 et al, the licensee has been -

required to " Provide an anlaysis and evaluation of the consequences of Appli- 1 l cable Condition II, III and IV Occurrences, in MODES 3 through 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 Instrumentation 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. l As part of this review, and as a safety.. justification.for'our concerns, we require inclusion of the following Occu'rrences and Considerations in the j program, and as early determinants of our proposals in respect of ECS Loop g ' Operability requirements in MODES 3, 4 and 5 (with loops filled)- G.2 DISCUSSION Item: CONSIDERATION j A number of factors determiJe our concern:

         'G.2.1        The increased boron concentration discussed under Section 3/4.1.1 of               .'

this review. G.2.1.1 Increases shut down margin at temperatures above 200 F, and thereby reduces the severity of any occurrences giving a return to power, - but only after reactor trip. Further the T.S. proposed by the licensee does not include the increased boron concentration and RCS Operability requirements are judged against those circumstances. G.2.1.2 Because increased shutdown margins are available, in MODES 3, 4 and 5, the licensee may now increase the level of withdrawal of all 1 movable control assemblies and still remain within the unchanged T.S. j condition of the allowable reactivity condition, keff of 5 0.99. . Consequently, it does 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 i 06/01/84 59 Revision A

to Trip. . Further, events giving a return to power after. reactor trip do not have improved initial protection; the reactor must still be - tripped prior to effecting the increased shut down margin ,and the-elimination of virtually all " Safety Related" levels of neutron flu) trip protection in TABLE 3.3-1 removes all current confidence'in -

                    "available" Reactor Trips on Neutron Power; the only Safety Related                   i Neutron Flux Trip from zero power subcritical conditions is the                       I Power Range Neutron Flux Low Set Point and the proposed T.S. removes                  !

this from operability in MODES 3, 4 and 5. Further it has a Safety d Analysis Limit of 35% power (25% Set Point) and together with related high peaking flux factors under these conditions is sufficient to g require all 4 RCPs running to ensure-R.C.S. Safety in at least MODE 3. G.2.1.3 The increased boron concentrations give less negative and more posi- , l tive moderate coefficients which changes the complexion ~and, nature of expected responses from " Licensing Bases Events." Under thtse cir-cumstances, it may not be-possible to validly deduce the resulting g responses and consequences without related analyses. ) 1 i G.2.1.4 At this time we see no protection against positive temperature

f. coefficients in MODE 3 [4, 5 & 6). Proposed T.S. page 3/4 1-4 ur 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 l all the rods withdrawn, end 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 i should also clarify this T.S. requirement which is ' l apparently in error and applicable to MODES 1 & 2 only , because.of the " RATED THERMAL POWER Condition." G.2.2 Removal of operability requirements for all safety related reactor trips (except SI) in Modes 3, 4 and 5, has placed the reactor in nonconformance with the requirements of 10 CFR Appendix A GDC 20,

                    " Protection System Functions" and GDC 22, " Protection' System Independence For All Occurrences Not Inititating Safety Injection."

Further, only a limited number of automatic trips (6) are blocked by existing plant permissive. P-7, 2 are blocked by P-8. This leaves

   ,                an additional 9 from which automatic protection can potentially be                  -

provided and which have been removed by unique action of the T.S. l without any Safety Evaluation. The proposed T.S. are nonconservative with respect to Regulatory 6 Requirements. They are also nonconservative in respect to the

 .                  Licensing Basis. The Licensee shail evaluate and propose.

G.2.3 InMODE3,downtoP-11,foreventsinitiatingSafetyInjeckion,the engineering within the existing Licensing Basis, might allow 10 CFR 50 Appendix A GDC 20 and 22 to be satisfied in respect to reactor trip and diversity. However, the proposed T S. does not propose 06/01/84 60 Revision A l

                                                                     +     -   _      -      -

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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 de is nonconservative. The licensee shall evaluate and propose. G.2.4 In MODE 3, f ro.n P-11, to MODE 5. for events initiating SI, the plant l is engineered and can be operated so that only one automatic trip of ~ the reactor may be available; that from containment pressure-high. On the above bases, plant engineering aw 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 projections on reactor trip may be available. In addition the proposed T.S.s do not require operability of the Reactor Trip /ESF channel 15 this phase of cperations below MODE 3 l [at P-11], to MODE 4 even though 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,8 asis and is I nonconservative. The Licensee shall evaluate and propose. g G.2.5 Diversity of Safety Injection to the maximum extent for related Accident Circumstances can only be retained within existing plant engineering by requiring that manual block of the Steam Line Pressure-Low be delayed until SG pressures are within an appropriate error band of the Steam Line Pressure-Low Set Point. This could be down to a temperature of approximately 485-490 F in the RCS which would be in MODE 3 before 1000 psig/425 F. (485-490 F is the satur-ation temperature equivalent to 565 psig + 30 psig [ channel error) 1.e., approximately 595 psig in the SG. l The licensee shall evaluate and propose.

l. -

m G.2.6 EVENTS OF CONCERN (A LIMITED SELECTION) G.2.6.1 OCCURRENCES WITH RAPID REACTIVITY INCREASE l Concerning " Uncontrolled Rod Cluster Control Assembly Bank Withdrawal from Sub-Critical Condition." ) Current Docketed Analysis in reference 7, section 15.2.1, pace 15.2-2-is based on four operating loops. This event is possible down to and including Mode 5.

Current FSAR analysis trips the reactor on Power Range, Neutron Flux-Low Set i l 06/01/84 61 Revision A l

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 mcderator reactivity coefficient is negligible (reference page 15.2-3, items 1 & 2). The event is initiated from hot 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.0PERABILITY of less than 4 RCS Loop in MODE 3 would be in nonconformance with the current FSAR

      )      in a nonconservative manner, and the licensee would be required to evaluate and l[$

i propose. Furthermore; increased boron concentrations would not change this requirement. -j f Additional events of a similar nature, with a rapid increase in reactivity ! include: ! a) Uncontrolled Boron Oilution (reference 7, pages 15.2-13) - G=)( As4e) , 4 b) Startup of an Inactive Reactor Coolant Loop (reference 7, page 15.2-19, - g) (,,;s e) revision 7) 4, .c ) Excessive Heat Removal Due to Feedwater System Malfunction (reference 7, ' gg page 15.2-30, revision 7) concerning initiation with the reactor at zero power). Until the licensee clarifies availability of MFW during MODES 3 'l through 5, this must be considered a potential occurrence. 'l 6 d) Single rod cluster control assembly withdrawal (reference 7, Page 15.3-9, * (h/USN revision 7). Although the Licensing Basis is at 100% power, the cir-cumstances from zero power should be reviewed. .

b. e) Rupture of a Control Rod Drive Mechanism Housing, at Zero Power (ref- -

  @M               erence 7, Page 15.4-30; revision 42).

f) Major Rupture of a Main Steam Line (see below). - G.2.6.2 STEAM LINE BREAKS: OCCURRENCES Concerning " Major Rupture of a Main Steamline" This event is discussed in Accident Analyses in Reference 7, section 15.4.2 and Reference 8 item 212.75 page Q 212-47d & e, item 25. Reference 8 proposes that the resulting impact on shutdown margins from this event during MODES 3, 4 and - 5 are improved over that of the design basis (of zero power, just critical, Tavg - 557 ) as:

                   " Operating Instructions require that the boron concentration be increased to at least the cold shutdown boron concentration before cooldown is initiated. This requirement insures a minimum of 1% ok/k shutdown margin at a Reactor Coolant System temperature of 200 F. This condition assures that the minimum shutdown margin experienced during the streamline rupture from zero power shown in the safety analysis is less than the case where safety injection 06/01/84                                 62                                                                                                                                   Revision A 1

s O , 6 actuation is manually blocked on low steamline pressure and low ) pressurizer pressure." 3 This position gives no measure of the resulting shutdown margins and/or power level and, the consequences of a stuck rod, with only 2 RC loops operating instead of four. It is conceivable 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 (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 MSLB cools the RCS down to 212 F e.g., the residual                   6 shutdown will be at 1% delta k/k whereas the proposed T.S. margin at Zero                @)(e Power according to T.S. Page 3/4 1-1_was 1.6 delta k/k. Please clarify, and at what condition during cooldown the 1.6% delta k/k is reached.
       . Given the circumstances that the " Operating Instructions" described above are               d>.

not a part of the proposed T.S., any T.S. allowing operability of less than @qsl 4 RCS Loops in M00E 3 would be in non-conformance 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) (above P-11) (reference W.0 0nl 1 i our required confirmation of this in an earlier item] so the Power Flux-Trip is not required to be Operable. At less than P-11, these circumstances are changed for the'MSLB, 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 related Reactor Trip System Instrumentation of any nature to Trip the Reactor and Insert the movable control rods to benefit from potentially increased available shutdown . margin. Ir. addition to all this, the licensee proposes that MSIV closure times under.these conditions in Not Applicable. Given the circumstances of the proposed T.S., and T.S. allowing OPERABILITY of 4

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, g% and the licensee would be required to evaluate and propose. Additional events which exhibit a rapid cooldown and depressurization of the RCS; are: a) Accidental Depressurization 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). 06/01/84 63 Revision A i

G.2.6.3 LOSS OF PRIMARY'C00LANT: OCCURRENCES Concerning: "Small Break LOCA" - This is discussed in reference 7, section 15.3.1 for a SBLOCA from rated power, and reference 8, item 212.75 page Q 212-47b for a 58LOCA between RCS conditions - of 1900 psig and 1000 psig/425 F in Hot Standby, and Q 212-64, item 3 together l 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 58LOCA 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 pumps through the core as they are coasting down following trip:      there-fore upward flow through the core is maintained."

Topical Report, 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 a) all pumps 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 from 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 pumps are assuaed to be tripped at the initialization of a postulated LOCA and a locked rotor pump resistance is used for reflood."

At this time therefore, the NRC must conclude that RCS pump operation and coast down is important to reducing the loss of core level subsequent to the event; also in maintaining unseparated two phase flow conditions and in ensuing rapid Boron (mixing and) Injection to the core. Rapid boron injection would not be an important issue if boron concentrations are already at cold shut down values, but minimizing loss of core level is important.

    ' G, until further evaluations are made, we must conclude that the current Safety                -

g Analysis Limits of the SBLOCA event is 4 RCS pumps OPERABLE in MODE 3 down to 425 psig/350 F. The current proposed T.S. are therefore non-conservative and (RSB) the licensee must evaluate and propose. Given the circumstances of the proposed T.S., operability of less than 4 RCS

 '     4 Loops in MODE 3 would be in non-conformance with the Current Safety Analyses Limits in a non-conservative manner and the licensee is required to evaluate N   and propose 06/01/84                                    64                               Revision A

Additional events of a similar nature to the SBLOCA events include: b a) Accidental Depressurization of the Reactor Coolant System (reference 7, page 15.2-33, revision 7). b) Steam Generator Tube Rupture (reference, page 15.4 - 13a, revision 38). I c) Rupture of a Control Rod Drive Mechanism Housing at Zero Power (reference 7, 6 page 15.4.6, revision 42). Both events, a) and b), are analyzed in the Licensing Bases at Full Power, and l 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, revis' ion 43, paras.1 and 5. Whereas both these l projections are proposed by the T.S. in MODE 2, they are not proposed for MODE 3 I which differs from the circumstances of MODE 2 by only a marginal reduction in J RCS Temperature. \ I J

The FSAR, reference 7, Table 15.4.6-1, revision 42, shows this occurrence i 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 respect 6-l to the Licensing Bases. The licensee shall evaluate and propose. W (834 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 LOCA between RCS conditions of 1900 psig and 1000 psig/425 F in Hot Standby 5; in item 212.90(6.3), page 212-61, for a LOCA at and 1.ess 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 4 RCS loop _ operation, 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 "Cnncerning Small Break LOCA" which used the W Topical Report WCAP 8356 (reference 19) and which applied to both Large and Small Break LOCAs. , l i l

06/01/84 65 Revision A ) l i

                                                                                 -______________________d

G m) Given the circumstances of the proposed T.S., any T.S. allowing OPERABILITY of less than 4 RCS Loop in MODE 3 would be in nonconformance with the Licensing Basis FSAR in a nonconservative manner, and the licensee is required to eval-uate and propose. G.2.6.4 OCCURRENCES CAUSING AN INITIAL INCREASE OF RCS TEMPERATURE 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: l 6 a) Main Rupture of a Main Feed Line (Reference 7, page 15.4-10, revision 30),

(g although this is normally evaluated at Rated power with no provision for evaluation as zero power.

(p513)

6. b) Start up of an Inactive Reactor Coolant Loop (W c) Loss of Offsite Power (reference 7, page 15.2-19, revision 7) 4.

 @HJd        d)    Partial Loss of Forced Reactor Coolant Flow (Reference 7, page 15.2-16, 6,              revision 7) y) R$8) 4        e)    Complete loss of Forced Reactor Coolant Flow (Reference 7, page 15.3-7, (w)(sq)          revision 7)

Except for item b; all these events are licensing bases events from Rated power, M 3 and not zero power, so that their importance would normally be minimal except M for the positive 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 prsitive temperature coefficients in MODE 3 (4, S & 6]. Given the circumstances of the proposed T.S., Operability of less than 4 RCS Loops in MODE 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. G.3 CONCLUSIONS Occurrence II, III and IV Events in MODES 3, 4 and 5, can result in returns to _ power with high peaking coefficients requiring effective reactivity control and/or reactor core flow for RCS protection, including DNBR, at the very substantially reduced pressure levels in the loop (2250 psig to 425 psig and less). Concomitant decreases in RCS temperatures are beneficial, but the importance of RCS pressure may be dominant. Acceptable RCS protection there-fore requires RCS flows which are substantial, and/or effective reactivity control including combined action to limit potential reactivity excursions. At this time, with the proposed T.S. , 4 RCS loops (with increased Reactor Trip Protection) would be required at entry into and during MODE 3 to meet the requirements of just the Licensing Basis Events From Zero Power. In MODE 4, 06/01/84 66 Revision A

operation of 4 RCS Loops, whilst on RHR, may be undesirable because of the substantial additional burden 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 Breaker.s, closure of Main Feedwater [ Containment] Isolation valves to both i Main and Auxiliary Feedwater Systems, Closure of Main Steam Isolation Valves, and Boration Control measures additional to those included in the proposed T.S. 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 plant down to effectively zero pressure (gauge) in the Steam Generators [or less if the condenser was still available] before transferring the heat sink-to the RHR system. This would ensure control of Steam Line Break, and LOCA                      l events, small and large, down to RCS conditions where RCS flows are not necessary.                                                                                         ;

Ot The current T.S. are nonconservative in respect to the Licensing Basis in respect to these concerns. The Licensee shall evaluate and propose. W T.S. SECTION 3/4.4.1: RCS LOOPS AND COOLANT CIRCULATION q START UP (MODE 2) AND POWER OPERATION (MODE 1). The LCO requires all [4] reactor coolant loops to be in operation in MODES 1 & 2. l The ACTION Statement 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 STANDBY 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 I 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 compensating condi-tions] would be non-conservative in respect of the existing FSAR. The Action Statement is' non-conservative with respect t'o the current licensing basis and the licensee shall evaluate and propose. T.S. surveillance requires verification of Reactor Coolant Loop (RCL) circula-t4cn once every 12 hours. This is unacceptable considering the Safety Analysis limits required above for loss at one pump. In the event of failure of the Lnw 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 consequences of a 10 min delay by the operator on hearing the Alarm - if it has remained operable from available [3 channel] LOGIC, Additionally, the FSAR proposes no alternate trips for the reactor, with related evaluation, such as over temperature leading to Pressurizer Level-High and Pressurizer Pressure-High. The Action Statement would place the plant outside the current licensing basis for normal operation and.is non-conservative with respect to that. The licensee shall evaluate and propose. 06/01/84 67 Revision A

l 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 - met: i

                   " Criterion 22- Protection system independence. The protection system'             -        i shall be designed to assure that the effects of natural phenomena, and of                   i normal operating, maintenance, testing, and postulated accident conditions                  ;

on redundant channels do not result in loss of the protection function' , or shall be demonstrated to be acceptable on some other defined 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." l The Facility is non-conservative-with respect to this Regulation, the licensee l

           .shall evaluate and propose. This is a generic issue.                                           l Thesurveillancerequirement,every12 hours,isintendedtoensurenotoniy                                 I that the system is operating, but that it is operating at process conditions which can be evaluated to show that the equipment is capable of performing its                      l Licensing Basis Safety Functions. The proposed T.S. requirements are absent                       .!

in this information; it is therefore non-conservative and the licensee shall i evaluate and propose. I T.S. Page 3/4 4-2: RCS HOT STAN0BY i 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 B 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 l least two loops be OPERABLE." 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; r.nd earlier work has shown the Safety Analysis Limits for the plant currently requiring at least 4 RCS pumps for this MODE. , l 4 The Action Statement allowing 72 hours with only one RCS loop operable is i non-conservative with respect to the current Safety Analysis Limits. -. At this time, any No. of loops less than 4 in MODE 3 is non-conservative with respect to the existing FSAR and the plant should be transferred to operation l in MODE 4 under these circumstances, with approved maximum normal cooldown rates. - ; I' 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 choice of such actions may be determined by " additional" protective procedures already in place at the plant, but not included in the T.S. where they are required by 10 CFR 50-36. Also, the particular combinations of projections 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. l 06/01/84 68 Revision A

'\

l Given the circumstances of the proposed T.S. , operability of less than 4 RCS loops in MODE 3, HOT STANDBY, would be in non-conformance with the current-Safety Analysis Limits in a non-conservative manner and the licensee is required -) to evaluate and propose, j It further follows, that the proposed surveillance requirement T. S. item 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 hours is also invalid and j should be changed. The surveillance requirement, once every 12 hours, is intended to ensure not j only that the system is operating, but that is is operating at process condi- ' tions which 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 information; it is therefore non-conservative and the licensee shall ; evaluate and propose. Surveillance requirements for the S.G. call for a level of 12% at least once f ', per 12 hours. Tnis is not in accordance with the Licensing Basis; this level-is the S.G. Low - Low Trip Set Point. All conditions II, III and IV occurrences ] 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 y alternate conditions. Therefore this exlisting proposal is outside the current i Licensing Basis and non-conservative. Reference our earlier comments under Item 2.1.1, Item f. The licensee shall evaluate and propose.

           *This Footnote proposes that;'in HOT STANDBY (MOUE ~3):
           "*All reactor coolant pumps may be de-energized for up to 1 hour 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 15.2-27, ;

           " Loss of Offsite Power to Station Auxiliaries"; but at MODE 2 Zero Power condi-tions with related programmed process conditions of Zero Load Pressure and Temperature in the loops. 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 comprehensive projections 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, isola-tion of MSIVs, and possibly an optimum boron concentration. At present, the i only Licensing Basis for controlling this particular situation is the Emergency 1 Operating Guidelines. I 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.

The licensee shall provide the reason for this requirement including the 1 expected condition of the Facility, and then analyze, evaluate and propose. l 06/01/84 69 Revision A i L__-__-_________-_______

Earlier concerns under General 2.6.1 addressed the need to evaluate the con-sequences of the Start Up'of an Inactive Reactor Coolant Loop in this MODE. No - apparent T.S. provision has been provided in the proposed T.S. The licensee shall evaluate and propose. 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 ACTION 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 proposal is non-conservative with respect to the Licensing Basis. The licensee shall evaluate and propose. T.S. Page 3/4 4-3. REACTOR COOLANT SYSTEM - HOT SHUTOOWN. ! The proposed T.S. should be supplemented by the conditions contained within the brackets [ ]:

       "3.4.1.3 At least two of the reactor coolant and/or residual heat removal (RHR) loops listed below shall be OPERA 8LE [and energized from separate power divisions] and at least one of the above reactor coolant and/or RHR loops shall be in operation:** [ Additionally two RCS loops must always be OPERABLE whenever RHR loops are in operation]

a. Reactor Cooiant Loop A and its associated steam generator [ including related auxiliary feedwater pumps] and reactor coolant pump,*

b. Reactor Coolant Loop 8 and its associated steam generator (including related auxiliary feedwater pumps) and reactor coolant pur,p.*

c. Reactor Coolant Loop C and its associated steam generator, [ including relating auxiliary feedwater pumps] and reactor coolant pump,*

d. Reactor Cnolant Loop D and its associated steam generator, [ including related auxiliary feedwater pumps] and reactor coolant pump,*

. e. RHR Loop A,*** and _

f. RHR Loop 8.***

APPLICABILITY: MODE 4. [Less than 425 psig/350 F]" g pg The licensee shall evaluate as outlined earlier under Item, General, for RCS 3 loops operability requirements and make proposals relative to the status of (R6dli 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, with the proposed TS in which limited boration is used and Reactor Trip System Safety Related Instrumentation and Safety Injection Instrumentation are all but 06/01/84 70 Revision A

4 l

I ' 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 energized from H.R A ) separate power divisions to protect against single failure of a bus or distri- y a bution system. When the RCS systems are used, the related Auxiliary Feedwater i systems are also required to be operable. The additional requirement proposed, for two RCS loops to be operable whenever RHR loop /s are in operation, is based upon reference 8, page Q 212-55 and 56, j 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:

                                                          "Approximately 3 hours are available to the operator to establish an alternate means of core cooling. This is the time it would take to heat                     ;

the available RCS volume from 350 F to the saturation temperature for 400 psi (445 F), assuming the maximum 24 hcurs decay heat load. To restore core cooling, the operator only has to return to heat removal l via the steam generators. The operator can employ either steam dump to ' the main condenser or to the atmosphere, with makeup to the steam genera- l tors from the auxiliary feedwater system. The time required to establish ! the a? ternate 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." ' MPA I The APPLICABILITY MODE 4, is necessarily qualified by [less than 425 psig/350 F] NgJ 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 l reference 8, page Q 212-47.d where it is described that l

                                                          "After several hours into the cooldown procedure (a minimum time is                          l approximately four hours) when the RCS pressure and temperature have                        {

l decreased to 400 psig and 350*F." And arising from a later revision 25, the FSAR advises on page Q 212-61b revi-sion 29 concerning ECCS calculations in a later submittal under Revision 28 that "The response provided in Revision 28 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 - has been assessed." , The additional Action statement that:

b. "With no reactor coolant or RHR loop in operation, suspend all operations involving a reduction in boron concentration of the Reactor Coolant System and immediately initiate corrective ACTION to return the required coolant loop to operation.

l l 06/01/84 71 Revision A A

i e and the additional notation that 1

                                                                                                                                     . i
                                              "***All   reactor coolant pumps and RHR rumps may be de-energized for up to 1 hour provided:      (1) no operations are permitted that would cause dilution of the Reactor Coolant System boron concentration, and (2) core outlet tempera-           -

ture is maintained at least 10 F below saturation temperature." are insupportable 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 STANOBY 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 i and propose.

Surveillance requirement 4.4.1.3.2 should verify S.G. water level at the Safety ) Mpp Analysis Limit for the Licensing Basis, which is the no-load programmed level, l

                                'g            not the current proposed TS value which is the S.G. Low-Low Level (Reactor                   j Trip] and AFW actuation. This proposed TS is non-conservative with respect                    1 to the current Safety Analysis Limits and the licensee shall evaluate and propose.

1 Surveillance requirement 4.4.1.3.3 verifying one loop in operation every 12 hours, j is insupportable as all protective trips on low flow in the RCP loops in this l condition have been removed. If low flow channel trips on the RCP loops are l 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, i page Q 212-56, item: j

                                                    " Case 1:    The Reactor Coolant System is closed and pressurized.                     )

I The operator would be alerted to the loss of RHR flow by the RHR low flow j alarm. (This alarm has been incorporated into the McGuire design)." Since currently, these two types of alarms are the only means of alerting the j operator to a Loss of Flow condition in the loop, which is beyond the Safety { Analysis Limits, then the alarms on both the RCS and Loop Flows should be j MPA Safety Related and included within the T.S.; and without further analysis at l g this time, two loops should be placed in operation. A proposal is made by the NRC for low flow alarms in each of the separated cooling systems, under Proposed T. S. Page 3/4 4-6a of this review. Regular surveillance should be proposed to ensure they remain operable as appropriate, over a specified surveillance period. - The Surveillance requirement, every 12 hours is intended to ensure not only that the system is operating, but that it is operating at process conditions g which can be evaluated to show that the equipment is capable of performing its design basis Safety Function. The current surveillance requirements for this biQh 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. 06/01/84 72 Revision A _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ c

4 Item 4.4.1.4.4 (Proposed). It is proposed that an additional item be inserted- ' which reads: "The related auxiliary Feedwater System shall.be determined (RSS OPERABLE as por the requirements of T.S. 3.7.1.2 [and 3.7.1.2.a as applicable)."

Current proposed T.S.s on T.S. page 3/4 7-4 are non-conservative in this matter by not providing any operability requirements for AFW in this MODE. The licensee shall evaluate and propose. An additional item is also required in which Atmospheric Dump Valves operability agpg is established. The current T.S. are non-conservative in this matter; they g{ggp 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 DUMP 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.] no The FSAR addresses the consequence of a failure, closed, of the isolation valve in the RCS/RHR line;-it addresses the analysis from 350 F in the RHR MODE when a bubble is present in the pressurizer. 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 in MODE 4, a loss of cooling could result in a potential overpres-surization of the system and the reviewer is not aware of any evaluation of the adequacy of the existing Low Temperature Overpressure Protection System to accommodate that event. The licensee shall evaluate and propose. T. S. Page 3/4 4-5: COLD SHUT 00WN (MODE 5] WITH LOOPS FILLED. The current proposed T.S. provides: ' 3.4.1.4.1 At least one residual heat removal (RHR) loop shall be OPERABLE and in operation *, and either:

a. One additional RHP loop shall be OPERABLE #, or

b. The secondary side water level of at least two steam generators shall be greater than 12%.

The current FSAR requires two (2) OPERABLE RHR tra' ins on two (2) redundant 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 necessarily l require S.G. temperatures greater than 212 F (Atmos Press in SGs) which would place it outside the Cold Shutdown MODE into the Hot Shutdown MODE - which is outside the required Functional MODE. The T.S. requirement for one RHR loop in operation and one to be available OPERABLE is currently not supportable by analysis evaluating the situation in which all RHR cooling is lost in a water solid condition; reference our 1 06/01/84 73 Revision A __m__ _ _ _ _ _ - - _ _ _ _ _ _ _ - - _ _ . - _ - - - - - - - - - -

5J ,

i

 -                immediately preceed'ing item T.S Page 3/4 4-3.                                                                                 In this case, if one only RHR loop is operating, loss of that single loop cause overheating in a water solidstate With potential overpressurization.                                                                                  Does the alarm of-loss of RHR Flow which is required, and an operator response time of.10 mins, provide                                                                                                  i sufficient time to commence operatior.s of the second RHR loop to'the extent                                                                                         .     !

neces:ary *to mitigate the consequences of.any potential overpressure event in an acceptable manner. The licensee shall evaluate and propose.

                -                                                                                                                                             3.

Use of secondary side water level of at le'ast 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 removing decay heat during Cold Shutdown. The proposed T.S. is therefore not in accordance with current Safety Analysis Limits, and. ) also non-conservative. As discussed in.the previous item T.S. Page 3/4 4-3, what'is required by the current Licensing Basis in Mode 5,.is to have available two OPERABLE RCS loops [ including AFW, SG and SG/PORVs] to meet the circumstances of failure closed o'f 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 LC0 reqc rement in the proposed T.S.~makes it non-conservative with respect 4 - to the Licensing Basis. The Licensee shall evaluate and propose. Footnote *: This item proposes that an only available operational RHR pump may be de-energized for up to 1 br. This event has not been evaluated, is not within the Licensing Basis, and is non-conservative. The licensee should (a define the circumstances, analyze and evaluate and propose. i The proposed surveillance requirement /4.4.1.4.1.2 pro'vides.that "At least one RHR loop shall be determined to be in operation and circulating reactor coolant at least once per 12 hours. The items of significance here are Operable Safety Related Flow Alarms with a surveillance frequency ensuring high probability of G, 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 hours, is intended to' ensure not only { that the system is operatirs La that it is operating at process conditions which can be evaluated to s w aat the equipment.is capable 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 G and propose.

T.S. Page 3/4 4-6. REACTOR COOLANT SYSTEM - COLD SHUTDOWN, LOOPS ARE NOT FILLED Item 3.4.1.4.2 requires that:

                   "3.4.1.4.2 Two residual heat remo/al (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 (2) redundant electrical buses so that each pump receives 06/01/84 74 Revision A

power from a different source; reference 20, pages 5.5-24, revision 9. Without this requirement, the T.S. is less conservative than the FSAR and the licensee shall evaluate and propose. 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 flow alarm.

Assuming worst case conditons (maximum 24 hours decay heat, air in the steam generator tubes, 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 off, if the operator took no action, boiling woLld 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 De limited 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 require 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 since boiling will have commenced. It is necessary to maintain two s operating RHR systems so that boiling may be eliminated on single failure. The licensee shall evaluate and propose. Additionally, the above information defines an LC0 of a minimum volume of water for the 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.5. as an LC0 with appropriate surveillance and Action Statements. A further T.S. require-ment is that any such min volume should be such that the level of water in or above the RCS loops be such as to provide acceptable flow, including NPSH conditions, over the range of temperatures expected, at inlet to the RHR pumps. Absent those required conditions from the Limiting Conditions of operation makes them non-conservative in respect to the Licensing Basis. 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 boron concentration of the Reactor Coolant System and immediately 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 operator would first attempt to fill and pressurize the reactor 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

c operator could-return ~the plant to heat removal via the steam generators. ' To do this the operator would have to jog the reactor coolant pumps to-sweep the trapped air from the~ steam generators. He would also have to open the steam dump valves (to atmosphere or.the main condenser) and ~ start up the auxiliary feedwater-system." In this MODE therefore, it is necessary to ensure that 2 RCS loops with operable 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 noncapability of restoring failed RHR systems to Operability within an acceptable time frame, including the possibility of. core uncovery in 100 mins. [The licensee shall also reference any Emergency Operating Guidelines in this respect]. Without provision for RCS, Loop Opera-bility required by the Licensing Basis FSAR, the current T.S. LCOs must be I considered non-conservative with respect to the Licensing Basis, and the licensee sball evaluate and propose. Item 4.4.1.4.2, A surveillance requirement, specifies: At least one RHR loop shall be determined to be in operation and circulating reactor coolant at least once per 12 hours. A time delay of 12 hours is excessive to verify a loop in operation, and this' has been considered earlier in this section. Further the surveillance require-ment, every 12 hours, 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 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 pump may be de energized for up to 1 bour provided:        (1)'no opera-                        I
            ' tions are permitted that would cause dilution of the Reactor Coolant System boron concentration, and (2) core outlet temperature is maintained at least                             'l 10 F below saturation temperature."

This departure from the Licensing Basis of two availa'ble RHRs with 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 minutes 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 Page 3/4 4-6(a) Proposed. A new subsection should be added entitled " REACTOR COOLANT SYSTEM, HOT SHUTOOWN TO REFUELING, APPLICABLE MODES 4, 5, & 6 which requires a LIMITING CONDITION OF OPERATION that two RHR Flow Alarms to Safety Related requirements shall be operable on each RHR loop when only one RHR loop is in operation 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 above. The current T.S. is non-conservative with respect to the Licensing Basis. T.S. SECTION 3/4.4.2 SAFETY VALVES SHUTDOWN (MODES 4 and 5} The T.S. requires that:

                                         "3.4.2.1 A minimum of one pressurizer Code safety valve shall be OPERABLE                   j with a lift setting of 2485 psig 2 1%.*

j APPLICABILITY: MODES 4 and 5. ACTION: ' L ' With no pressurizer Code safety valve OPERABLE, 'immediately suspend all operations involving positive reactivity changes and place an OPERABLE RHR loop into operation in the shutdown cooling MODE." 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 j Licensee shall evaluate and propose. ' The Action statement is based (reference T.S. page B 3/4.4-2) on the premise ! that INOPERABILITY of the Safety Valve in Modes 4 and 5 needs to be offset by operability of pressure relief. valves in the RHR systems. This is not the safety basis for Action. The safety basis is, that the Reactor Coolant Pres-sure Boundary has been effectively rendered inoperable requiring the operator to proceed to a cold shutdown condition with the zero pressure (gauge) in both RCS and SG systems, and related reactivity control actions to ensure that no { return to nuclear power is possible. This needs to be done in a manner I consistent with the nature of inoperability of the Safety Valve. The current T. S. is nonconservative with respect to the Licensing Basis; the licensee shall evaluate and propose. Further, 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 l and contrary to the Licensing Basis. The licensee shall evaluate and propose. The Surveillance Requirements should contain the minimum discharge capacity 6q required of this valve as defined in the Licensing Basis. They should also gggg ensure the maintenance of satisfactory environmental conditions consistent with reliable valve operability. The licensee shall evaluate and propose.

                                                                                                                         .a l

06/01/84 77 Revision A

m a . T S. Section 3/4 4.2 SAFETY VALVES' OPERATING The proposed T.S. requires all [3] pressurizer Code Safety Valves to be Operable in Applicable Modes 1, 2 and 3. The Action Statement requires that

        " ACTION:                                                                                  .)   i With one pressurizer Code Safety Valve inoperable, either restore the inoperable valve to OPERABLE status within 15 minutes or be in at least HOT STANDBY within 6 hours.and in at least HOT SHUTDOWN within the following 6 hours." l k.

q Failure.of the Pressurizer Code Safety Valve, in' general, would infringe the ' integrity of the Reactor Coolant Pressure Boundary and the RCS should be brought 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 inoperability, and potential for all positive reactivity levels eliminated. The worst situation would be that of an " Accidental Depressurization of the l Reactor " Coolant System" analyzed for the most severe conditions' including

j maximum core power, reference 7, page 15.2-33' revision-7. This type of event would require Emergency Procedures to define the ACTION STATEMENT. 4 1 l Could other types of Vailure allow other-types of response which could be- l outside the Emergency.0perating Procedures. The Licensee had not identified - others and analyzed and evaluated the related safety to Regulatory Require-ments as a basis for his proposed action, i The T.S. Bases on page B 3/4.4-2 does not exhibit an acceptable understanding of the importance of,.and potential severity of, the event. including failure k types and appropriate Regulatory requirements including p.rocedures.

       -The existing ACTION statement is inadequate within the Licensing Basis, and therefore unacceptable. The only existing Licensing Basis must be within the analyses reported in reference 7. page 15.2-33, revision 7, and the proposed j

Action Statement does not recognize these circumstances. The existing Action l Statement is therefore nonconservative with respect to the Licensing Basis; j the licensee shall evaluate and propose.

                                                                                              -1 LCO and surveillance procedures must also address position indication and/or discharge flow measurement procedures, including pressurizer relief tank condi-                )

tion and other measures to ascertain the operability of the valve [this is l necessary to satisfy 10 CFR 50 Appendix A, Criterion 20, 32 and 33). The writer reviewed, in 1983, information pertaining to the GPU/B&W lawsuit review, I and his recollection is that the TMI-2 operators " initially thought tha*. the ; safety valves had developed a leak in.the PORV3 because the valves had lifted on a recent event." There must be a measure of acceptable leak tightness from 06/01/84 78 Revision A

l L measurable parameters "in operation" tc ascertain the status of the valve so that acceptable measures can be taken. The safety basis for the concern' rests not only in the previous position addressed above, but also, that in the event of failure of control grade " pres-sure control devices" these valves will be challenged on the-following occur- ; rences within the Licensing Basis. J Startup of the Inactive Coolant 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 Break Accident, reference 7, Figure 15.4.2.7, revision 38-One Locked Rotor Event; reference 7, Figure 15.4.4-1,-revision 32 J l i l Safety Valve Operation could also occur on other overpressurization events if ' same of the early retctor trips fail to operate as expected.

In this matter, the T.S. is nonconservative with respect to Regulatory Require- 1 ments. The Licensee shall evaluate and propose. This could be a generic issue. l Surveillance Requirements should reference the documents containing the record

of the Inservice Testing of the valves for inspection on a regular basis'of 12 hours so that changing operating staff are kept aware of a potentially 1 changing status on a singularly critical item. }

1 T.S. Section 3/4.4.3 PRESSURIZER l l T.S. Page 3/4 4-9 The APPLICABILITY MODES are proposed as 1, 2 and 3. Item: Pressurizer Level: The response of all the analyses of Condition II, III and IV events in refer-ences 7 and 8 depend upon an initial level of water in the Pressurizer.which is programmed as a varying value dependent upon.the Nuclear Power Level'. Addi- - tionally, the response of all Condition I events which determine the most 1 conservative set of parameters from which to start Condition II, III and IV 1 events, are also so dependent upon this same programmed pressurizer level. Since therefore this pressurizer level is used in establishing an teceptable 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 allowable values consistent with the related g, channel errors and Safety Analysis Limits used in the FSAR, Section 15 in reference 7. The licensee shall evaluate and propose. [II 06/01/84 79 Revision A , o I

l l

   .M      APPLICABILITY MODES:      Pressurizer level should be proposed for MODES 1, 2, 3, (Rss) and 4 (with steam bubble). Down to MODE 4 is provided to cover LOCA and                   ,

MSLB events considered in reference 8. Also, the plant can then be placed on Automatic Level Control. Appropriate ACTION and SURVEILLANCE procedures should be proposed. Licensee shall evaluate and propose. Item: Pressurizer Pressure The responses of all the analyses of Condition II, III and IV events in refer-ences 7 and 8 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 determine the most conservative ; set of parameters from which to start Condition II, III and IV events, are also 1 so dependent upon this same pressurize pressure. 1 Since therefore this value of pressurizer pressure is used in establishing an j acceptable 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, for each of MODES 1 through 5, the ] licensee should provide details of the pressurizer pressure Set points with s allowable values consistent with the related channel errors and Safety Analysis Q Limits used in the Licensing Basis in the FSAR in Section 15 in reference 7, (eds and reference 8. The licensee shall evaluate and propose. I 6 Appropriate ACTION and SURVEILLANCE procedures should be proposed. The licensee gg) shall evaluate and propose, , T.S. SECTION 3/4'.4.4 RELIEF VALVES (POWER OPERATED) The current T.S. provides that the plant may continue in operation if either j one of the combination of Block Valve and PORV is INOPERABLE. This is a i contravention of the regulations which provides under 10 CFR 50.2(v) that: i (v)" Reactor coolant pressure boundary" means all those pressure-containing 1 components of boiling and pressurized water-cooled nuclear power reactors, such as pressure vessels, piping, pumps, and valves which are:

               , (1) Part of the reactor coolant system, or (2) Connected to the reactor coolant system, up to and including any and all of the following:
                                                                                                         ~

(i) The outermost containment isolation valve in system piping which penetrates primary reactor containment. (ii) The second of two valves normally closed during normal reactor operation in system piping which does not penetrate primary reactor containment. (iii) The reactor coolant system safety and relief valves. Since a single failure of either the Block valve, or the PORV, will reduce the level of protection of the Reactor Coolant Pressure Boundary (RCPB) from two 06/01/84 80 Revision A l

(2) valves to one (1) only valve, the Regulatory Requirements are not met and the plant must proceed to a cold shutdown condition with no potential for positive reactivity cnanges, within appropriate time frames. The current T.S. is nonconservative in respect to' Regulatory Requirements.

   ]                           The licensee shall evaluate and propose.

T.S. Section 3/4 4.5 STEAM GENERATORS T.S. Page 3/4 4-11 a) 5.G. Levels b A number of the Accident Analyses.in reference 7 depend upon an initial level b of water in the Steam Generator. A specific example is the Main Feedwater Line Rupture 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 occurs; reference related Table 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 levels including related Safety Analysis Limits, and respond to the proposition that such values should be included as Set Point values and Allowable values in the proposed T.S. as Limiting Conditions of Operation for the facility with appropriate Action Statements. The proposed T.S. is nonconservative by their

absence.

                                                                                                                                                   """~'

4 b) Steam Generator Pressures l Since Steam Generator Pressures and related Saturation Temperatures under i normal steady state operation can be a significant determinant of system ' responses for Condition II through IV occurrences analyzed in the Licensing Basis including Section 15 of reference 7, and reference 8, please provide the values used as Safety Analysis Limits in related analyses and again respond to the proposition that such values should be included as set Point and Allowable values as Limiting Conditions of Operation for the facility with appropriate Action Statements. The proposed T.S. ,is nonconservative with respect to the Licensing Basis, by their absence. _ c) Please respond to the proposition that this section should also adequately identify the maximum allowable Steam Generator Pressure under Transient and _ Accident conditions with appropriate Action Statements. Maximum SG pressure j is one of the Acceptance Criteria for safety. The current very limited basis i for Steam Generator Pressure integrity is completely inadequate. Please l 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 and the value quoted in the T.5. Basis Page B 3/4 7-1 at 1185 psig. The proposed T.S. is nonconservative with respect to the Licensing Basis, by this absence. 06/01/84 81 Revision A

d) APPLICABILITY MODES 1, 2, 3, and 4: The current applicability requirements relate to Structural Integrity RMS considerations. On inclusion of Steam Generator Level and Pressure as determinants of Opera-bility, the licensee should evaluate and propose APPLICABILITY MODES consistent with RCS/SG loop requirements discussed in this review under separate sections and particularly under Reactor Coolant System and Residual Heat Removal sections in MODES 1 through 5. This will embrace operability requirements from MODES 1, 2, 3 and 4 through 5. The proposed T.S. is nonconservative with respect to the Licensing Basis, by the absence of this information. The licensee shall evaluate and propose. T.S. Page 3/4 4-36 (REACTOR COOLANT SYSTEM) OVERPRESSURE PROTECTION SYSTEMS The current LCOs require that either of the following 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 when the temperature 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 the 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 pressurizer water volume is less than 1600 cubic feet, or (2) the secondary water temperature of each steam generator is less than 50F above each of the Reactor Coolant System cold leg temperatures. It is necessary, to expand the LCOs to all those which should be incorporated into the operability requirements for the pressurizer and steam generator dis-cussed earlier under T.S. Section 3/4.4.3 Pressurizer and T.S. Section 3/4.4.5 Steam Generators. This additional information defines necessary safety limits for the Licensing Basis event; as in reference 28, which is an early Topical Report submitted by W for approval. The proposed T.S. is nonconservative in 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 graater than or equal to 4.5 square inches: We find that this should be confined only to MODE 5, COLD SHUTDOWN, LOOPS ARE N01 FILLED, and REFUELING OPERATIONS; MODE 6 HIGH WATER LEVEL and MODE 6 LOW WATER LEVEL. There are no safety analyses to support this type of operation in remaining MODES 4 and 5. The proposed TS, without this clarification, is nonconservative with respect to the Licensing Basis. The licensee shall evaluate and propose. 06/01/84 82 Revision A

       .g 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 proposed T.S. The licensee shall evaluate and propose.

l I l

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I 1 I 06/01/84 83 Revision A f I \ \ .

T.S. SECTION 3/4.5 EMERGENCY CORE COOLING SYSTEMS The operability requirements from the McGuire Units 1 & 2 Licensing Basis FSAR j are markedly different from those of the W Standard Technical Specifications ' which have been adopted by the Licensee in his proposed T.S. The Licensing Basis FSAR requirements are summarized under " General." General i i 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 except they will be in reverse order. The startup procedures are not presented here to avoid unnecessary duplication.

I Operator Instructions During 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 pressurizer 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, locks and tags the breakers for these ' valves. He also opens locks and tags the breakers for all safety injection pumps and all but one charging pump. At this time, one charging pump and two residual heat removal (RHR) pumps would be available for either automatic or manual SI actuation. C) At less than 400 psig and 350 F, the operator aligns the Residual Heat Removal System. The valves in the line from the RWST are closed. II Operator 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 l the operator. i 1 l l l 06/01/84 83A Revision A l l

4 , , M 8) Between 1000'psig.and 400 psig, a portion.of the ECCS can be actuated automatically-(containment high pressure signal) or manually:by the . 1 operator, 'The equipment-that can be energized are two RHR pumps'and on,e charging pump. The operator would have to reinstitute power at- l l the motor control centers or switchgear_to the remaining safety _ e ;

                -injection pumps, charging pump, and the accumulator isolation valves.

j y g C) Below 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." - In* response to additional questions, the following information was provided under.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. ,

                                                                                                                                                           .l
                                                                                                                                                           'l
            .The above plant operating procedures will ensure that the accumulator I.           isolation valves will not be locked out prior to about 2-1/2 hours af ter                                                                        (
                                                                                                                                                            ;I I

reactor shutdown for a cooldown rate of 50*F/hr. A conservative analysis has defined that the peak clad temperature 3 resulting from a large break LOCA would be significantly less than the. 2200 F Acceptance Criteria limit using the ECCS equipment available 2-1/2 hours after reactor shutdown. j! l The following assumptions were used in the analysis: l 1

1. The RCS fluid is isothermal at a temperature of 425 F and a pressure _

d of 1000 psig. ;

2. The core and metal sensible heat above 425 F has been removed.

l

3. The hot spot occurs at the core midplane.

4. The peak fuel heat generation during full power cperation of 12.88 kW/ft (102% of 12.63 kW/ft) will be used to calculate adiabatic heatup.

5. At 2-1/2 hours decay heat in conformance with Appendix K of 10 CFR 50, i' the peak heat' generation rate is 0.179 kW/ft.

06/01/84 84 Revision A

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                   .a.                                                                                                                                                                  j

6. Two' low head safety injection pumps and'one high head charging pump.

a are available from either manual ' Safety Injection actuation or , automatic actuation by the containment Hi-1 signal.

       .                                                                                 7. No liquid water is present in the reahtor vessel at the end of-blowdown.

8. A large cold l'eg break is considered. l 1

For a postulated LOCA at the cooldown condition of 1000 psig, pre'vious calculations show that the clad.does not heat up above its initial temperature during blowdown. Proceeding:from the end of blowdown and assuming adiabatic heatup 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 generation in the core becomes sufficient to inhibit the reflooding rate. At that time, heat transfer from the clad'at the l hot spot.to the steam boiloff and entrained water will commence. -This l ' heat 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. Downcomer 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 when the core, water level reaches the core midplane,

2. Core reflooding is delayed until the 51 pumps have completely filled the downcomer. No cooling.due to steam boiloff is considered at the. !

l 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 temperature rise thereafter. For Case 1 described above, the water level reached the core midplane 43.2 seconds after bottom of core recovery,. The temperature rise during-reflood at the hot spot from adiabatic 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-ponding temperature rise at the hot spot form adiabatic heatup is 272 F; which gives a hot 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 shutdown case. The combination of more rapid reflooding and lower power in the fuel insures that the clad temperature rise during reflood will be less for the shutdown case than for the design basis case. l L 06/01/34 85 Revision A l

    ~                                                     =

c Repeating-the'above calculation. assuming the loss.of a low' head safety injection pump yieldsaclad temperature of 1653 F and 1760 F.for Cases,1 and 2, respectively. Thest results provide additional assuran:e that .the ' peak clad temperature will not exceed 1130*F because, as stated above, 'iri-the shutdown case more rapid.refloodire; and lower power in the fuel'

insures that the clad temperature rise caring 'reflood wil be less. than for the design basis case.

Based upon the analysis as presented (cove, 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.

                                                 ~
                                                                                                               ~

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 has been assessed. The initial conditions would be reached l four hours ~after reactor shutdown. The integrity of'the core after a postulated LC0A is assuredfif 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-l 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: 1 (a) The system pressure essentially reaches: equilibrium with containment by the t.ime 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 fluid considered. l (c) Volume between the core barrel and baffle is conservatively neglected. (d) 120% of the ANS decay heat curve for four hours after shutdown is utilized. Using the void fractions developed from the Yeh correlations and utilizing a hydrostatic pressure balance, the height of the steam-water mixture in

  -           the upper plenum was generated. Incorporating the plant geometry, the                         _ 3 total liquid mass in the downcomer, core, and upper plenum was calculated,                             !

i.e., a mass-initial condition. Again by hydrostatic p' essure balance, i the height of liquid in the downcomer when the top of tha core is just about to uncover was calculated. This information along with core volume is used to develop a mass-final condition. That is, the mass is liquid 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 compared to the ten minute assumption for operator reaction time. ] I 06/01/84 86 Revision A l l-A , ,

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-time assumption."

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." Additionally, the status of the ECCS systems from entry into the RHR MODE through cooldown, i.e., from 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 through both MODES 4 and 5:

              "5.2.2   Overpressure 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 vehen the reactor coolant system is water-solid. It consists of two separate trains each containing a
 .           power operated relief valve set to open when the system pressure reaches 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 enabling of the water-solid overpressure protection system; gnabling is performed manually, by turning key-lock 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, each train contains an annuciator which . sounds when the power-operated relief valve is open, indicating an overpressure transient is in process.

Each power-operated relief valve is supplied with nitrogen from the' cold leg accumulators. No operator action is required in the event of a transient. The operator isolates the upper head injection system, the cold leg accumulators, the safety injection pumps and one centrifugal charging pump before the 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-tent start with concomitant mass addition. The only other overpressure event _. would result from an inadvertent main coolant pump start with the coolant in . the secondary side of the steam generator hotter than that in the reactor I coolant system. The applicant has shown that in neither case was 10 CFR Part 50, Appendix G limit reached. For the'latter case (that for main coolant pump inadvertent start), the applicant assumed that the temperature of the fluid in , the steam generator would exceed that in the reactor coolant system by no I greater than 50 dagrees Fahrenheit. The staff requires that the technical specifications require that the reactor coolant system may not be cooled to temperatures lower than 300 degrees Fahren-heit without the overpressure protection system enabled, and unless both 06/01/84 87 Revision A I

q 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' o 1 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 . j Fahrenheit when the reactor coolant system fluid temperature is less than j

            ,                                 300 degrees Fahrenheit since the applicant's calculations did not assume                                 1 differences-greater than 50 degrees Fahrenheit.

The applicant provide'd' data to show that the power-operated relief valve opens { within the time specified in the analyses. I I The system meets the single failure criteria as only one of the two trains 2is 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 j and Electronics Engineers Standard 279-1971, " Criteria for Protection Systems." This system meets the staff requirements for an overpressure protection system j with the reactor coolant system water-solid and is acceptable. We consider this matter resolved? -] j 1 The required status of the ECCS systems required by the existing Licensing -l Basis FSAR are briefly summarized: 1 Above 1900 psig (in MODES 1, 2, and 3): All ECCS systems are OPERABLE. 1 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 . i closed and breakers racked out,1 ceritrifugal and 1 reciprocating charging l pump and.2 safety injection pumps are isolated, and rendered inoperable by 1 opening and locking the related circuit breakers. Below 425 psig/350 (in ) MODES 4 and 5) status of all ECCS systems remain unchanged, i.e., same (astfor -] the preceding phase of MODE 3) with the exception that remaining equipment is j 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 pumps are ! effectively electrically isolated.] RHR PORVs are rendered operable 'during , water solid operation, below 300 F. 1 These requirements are substantially different from those of the W STS which j the licensee has adopted for his facility contrary to his Licensing Basis as 1 disclosed in the FSAR and SER to the above references. _ 1 I T.S. SECTION 3/4 5.1 ACCUMULATORS / COLD LEG INJECTION l Item: APPLICABILITY MODE

    @ss)                                       The Applicability Mode, given as MODES 1, 2 and 3* where 3* is 1000 psig, should be amended to include 425 F; as 1000 psig/425 F. Reference the basis in the previous section entitled " General."

Since the proposed T.S. does not contain this temperature constraint, it is non-conservative. A pressure of 1000 psig on the current Appendix G curve, 06/01/84 88 Revision A

and T.S. temperature constraints, would permit an RCS temp of 557 F. The only available analysis in the Licensing Basis, see earlier unde " General," shows that cooling down to [1000 psig]/425 F is necessary 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 28. The current T.S. is therefore non-conservative in this matter, and the licensee must evaluate and propose. Note; the " Footnote

Pressurizer Pressure above 1000 psig" also needs amendment.

Item: 3.5.1.1.d. Nitrogen cover pressure is quoted at between 400 and 454 psig. The Licensing Basis FSAR, reference 4, page 1 of 5 revicion 39 in Table 6.3.2-1 specifies a normal operating pressure of 427 psig. Making an allowance for channel error and drift should not this value be a higher set point of approx. 450 psig. The specified set point values proposed in the T.S. of 400 to 454 psig can therefore give actual values which are lower than in the Licensing Basis FSAR and be non-conservative. The Licensee shall evaluate and propose. Item 3.5.1.1.f Proposed The NRC proposes that an additional item limiting the range of actual water temperature in the accumulator between 60-150 F in accordance with Licensing Basis FSAR reference 29, Table 6.3.2-1 is necessary to confirm Safety Analysis Limits for this accumulator. Its absence from the proposed T.S. renders 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. 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 whether HOT SHUTDOWN is necessary instead of to 1000 psig/425 F. Further, is there any conservative direction of the error which may minimize his need to suspend operations at power, or allow him to operate at reduced levels. This licensee proposal may be necessarily conservative. The licensee may evaluate and propose. Item 4.5.1.1.c requires that "once per 31 days when the RCS pressure is above 2000 psig, it is verified that power to the isolation valve on the Cold 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 once per 18 months verify that each accumulator isolation valve opens automatically under each of the following conditions:

1) When an actual or a simulated RCS pressure signal exceeds the P-11 (Pressurizer Pressure Block of Safety Injection) Setpoint,"

We are not aware that this actually occurs; the licensee shall review and advise of the related details with'in the FSAR on other licensing basis records'. This action is not described in FSAR reference 7, under Table 7.3.1-3 (1 of 2) 06/01/84 89 Revision A I

v - and (2 of 2) revision-35, " Interlocks.for-ESFAS," nor in the related Logic Diagrams. , The LCOs of tne Licensing Basis FSAR require that this Cold Leg Injection Accumulator be m'de a operable whenever 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 which 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 evaluate and propose.

The licensee sha'll verify that the set points for the-relief valve on the Accumulators are included in the Inservice Testing Program at the facility. T.S. Section 3/4.5.1.a (Proposed) An additional T.S. Section is proposed that provides specifically for the fact

that " COLD LEG INJECTION ACCUMULATOR ISOLATION VALVES" at " APPLICABLE CONDI-TIONS" of MODE 3 (< 1000 psig/425 F), MODE 4 and MODE 5 would have a " LIMITING CONDITION OF OPERATION" providing that "Each Cold Leg Injection Accumulator Isolation Valve _is closed with circuit breakers opened, locked and tagged." Appropriate Action Statements and Surveillance Procedures would be provided. This is in accord with the LCOs of the Licensing Basis FSAR as described under , earlier items T.S. 3/4.5, " General" and T.S. 3/4.5.1 of this review. . Absence i of this specific provision makes the proposed T.S. non-conservative The I licensee shall evaluate and propose. . T.S. Page 3/4 5-3. UPPER HEAD INJECTION l Item: APPLICABILITY MODE. l The Applicability Mode given as MODES 1, 2, and 3* where a signifies Pressurizer i Pressure above 1900 psig, should be amended to-include >425 F; as 1900 psig/>425 F. The FSAR does not include the temperature constraint explicitly at 1900 psig, though it is implicit in that the next lower. boundary for change is 1000 psig/425 F [ Reference earlier Item: T.S. 3/4.5 under GENERAL]. Absent this condition, the related 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, includes overpressure concerns at the reducing levels of temperature down to 425 F, reference 12, page 7-1. From his detailed analysis, the licensee should evaluate and propose a lower limit - to this teraperature condition of >425 F. Item 3.5.1.2.c Nitrogen cover pressure is specified as between 1206 and 1264 psig. The Licensing Basis FSAR, reference 29, page (1 of 5), revision 39 in Table 6.3.2-1 specifies a normal operating pressure of 1220-1280 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 proposed T.S. of 1206 would therefore require lower pressure in the RCS before actuation and is therefore non-conservative. The licensee shall evaluate and propose. l 06/01/84 90 Revision A

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), revision 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 STANDBY, generally, except for closed isolation valves, followed by HOT SHUTDOWN. This may be too conservative - the licensee should review specifically each of the Operability items b, c and proposed d, and decide whether HOT STANDBY leading ultimately to HOT SHUTDOWN is necessary. Further, he should assess if either boundary value, upper or lower, can be conservative, and by how much, and evaluate whether he should take an ACTION STATEMENT under " conservative" conditions. The licensee may evaluate and propose. The licensee shall verify that the relief valve set point on the Accumulator is included in the In Service Testing Program at the facility. T.S. Section 3/4.5.1.b (Proposed) (RSB) An additional T.S. item is proposed that provides specifically for the fact that " UPPER HEAD INJECTION SYSTEM ISOLATION VALVES" at APPLICABLE CONDITIONS of MODE 3 ($ 1900 psig and > 425 F), MODE 4 and MODE 5, would have a " LIMITING CONDITION 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 accordance 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 specific provision makes the current T.S. non-conservative with respect to the Licensing Casis. The licensee shall evaluate and propose. T. S. Section 3/4.5.2 ECC SUBSYSTEMS -Tavg 1 350 F The title should be amended to read as: ECCS SUBSYSTEMS - PRESSURIZER PRESSURE > 1000 psig/RCS Tavg34 25 F - The Operability requirements of 2 full trains of ECCS equipment remains unchanged. Absence of the pressure / temperature condition in the proposed T.S. is not in accordance with Safety Analysis Limits. Its absence permits high pressure pump operation at lower pressures and temperatures with potential infringement of related safety criteria. Related safety criteria have not been well defined, or docketed, but are apparently considerations of Low Temperature Overpressure Protection of the RCS under these and related Accident circumstances including inadvertent operation of ECCS pumps. This diversion from the Safety Analysis I 06/01/84 91 Revision A

Limits of- the Licensing Basis FSAR must therefore be ' considered non-conservative' .. and the licenseee shall evaluate and propose. Item 4.5.2.h.: concerning flow balance tests'in the ECCS system. .The licensee ' shall provide the bases. for the flow distributions specified 'and further ~advire how they might meet. minimum flow conditions-to intact loops-dating Accident

Occurrences.

l. T.S. Section 3/4.5.2.A Preposed A proposed new Section which would be titled: =ECCS Subsystem - Appli.cability between 1000 psig/425*F.and 425 psig/350 F.

!_ This would provide for: One ECCS subsystem comprising the'following shall be j -OPERABLE-l

a. One OPERABLE centrifugal charging pump,#

l-

b. One OPERABLE RHR heat exchanger, l l

c, One OPERABLE RHR pump, and j i

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 breakers for all safety injection pumps and.all but the one operable i centrifugal charging pump are opened, locked and tagged (reference earlier information). l As explained in the previous section, limited operation of the higher pressure pumps between 1000 psig/425 F and 425 psig/350 F apparently provides Low Temperature Overpressure Protection (L70P). The proposed T.S. requires all CI and SI pumps to be available during these conditons and is therefore j non. conservative with respect to the Licensing Basis and particularly in respect 4 of Overpressure Protection. The licensee shall evaluate and propose, and in so- - doing provide the analyses and evaluation which required constrained operability of the higher pressure pumps in this operating phase, in his Licensing Basis FSAR. T.S. Section 3/4.5.3 ECCS Subsystem - Tava 1 350 F This title should be amended to read ECCS Subsystems -.425 psig/350 F to COLD SHUT 00WN The current T.S. provides 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 06/01/84 92 Revision A I I

RCS pressures" of 2485 psig under these circumstances. Also the proposed T.S. alignment eliminates safety injection and charging pump capacity. There is no available evaluation of the capability of the reduced ECCS system to satisfac-

torily mitigate the consequences 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 Basis 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 SHUTDOWN 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. Reference 8 pages Q212-56 revision 25 under the Titles of Case 1 and Case 2 and page Q 212-57, revision 25, under the Title of Case 3. Overpressure Protection also, which 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 comprisirg

a. One Vperable Centrifugal Charging Pump (CCP)

b. One Operable RHR Heat Exchanger

c. One Operable RHR Pump

d. An Operable Flow Path There are no Safety Analyses or Evaluations of one only ECCS subsystem allowing for a single active failure in one only train. This proposition is therefore non-conservative with respect to the Licensing Basis FSAR. The Licensee shall evaluate and propose.

This T.S. does not disallow the additional CCP and 2 Safety Injection Pumps (SIPS) from 350*F down to 300*. This again is non conservative with respect to the LCOs of the Licensing Basis FSAR which allows only one (1) CCP, and the remainder i.e., one (1) CCP and any other reciprocating charging pump and 2 SIPS are to be electrically isolated against inadvertent operation. This proposed T.S. is again non-conservative in respect of overpressure protection when com-pared with the current Licensing Basis. The licensee shall evaluate and 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 OVERPRESSURE 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 licensee shall evaluate and propose. The LCOs of the Licensing Basis FSAR require the same operability of ECCS equipment as is required for TS 3/4 5.2A Proposed. So that in addition to: 1 06/01/84 93 Revision A

(.

     .m One ECCS subsystem comprising 'the following shall. be OPERABLE:

a. One OfERABLE centrifugal charging pump,

b. One OPtiRARLE RHR heat exchanger, *

c. One OPERABLE'RHR pump, and ,

i

d. An OPERABLE flow path ~

l which is the same as for the proposed T.S.,.it is also required that: One ECCS subsystam comp,-ising the,following shall also be' OPERABLE: 1

b. One OPERABLE RHR heat exchanger,

c. One OPERABLE RHR pump., and s j

                   .d. An OPERABLE flow path.

Additionally, that'all breakers for all safety injection pumps and all but the one operable centrifugal charging pump are opened, locked and tagged. , I (reference earlier information) The proposed T.S. is therefore less conserva- ' tive than the Licensing Basis FSAR by being deficient in ECCS total pumping capacity, and excessive in available high pressure pumping capacity so infringing LTOP. The licensee shall evaluate and propose. I Additionally the Licensing Basis requires' that ech of these subsystems be j independent and receive power from two (2) redundant Emergency Buses and 1

  -,         Power Sources.      The absence of any such provision in the proposed T.S. makes                1 it non-conservative with respect to the Licensing Basis. The Licensee
          , shall evaluate and propose.

T/S Section 3/4.5.4 BORON INJECTION SYSTEM / BORON INJECTION TANK. 6: Item: APPLICA81LTY MODES 1, 2, and 3 with the current proposed T.S. should'be changed to include MODE 4 in accordance with the Licensing Basis FSAR which evaluates MSLB and LOCA events down to and. including this_ MODE. Adoption L 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 be considered non-conservative. The licensee should evaluate and propose. Item: The ACTION Statement should be clarified to include [ ] that in the event of.inoperablity of the BIT tank, the RCS be borated to [a boron concentra-tion which will give] a SHUTOOWN margin of 1% delta k/k at 200 F. The licensee shall clearly indicate, that this item is not applicable to Unit 2 by reason of a recent SER from NRC. Comment: Since BIT concentrations of only 2000 ppm, only are now required, 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 SHUTDOWN overly conservative; if minimum volumetric requirements are necessary, can 06/01/84 94 Revision A L o _

additional provision be made in the RWST. The licensee may evaluate and

   ,                 propose.

T.S. Section 3/4.5.5 REFUELING WATER STORAGE TANK Item: APPLICABILITY MODES 1, 2, 3, 4. The. current MODES 1, 2, 3 and 4 which includes an LCD for 372,100 gallons must be extended to MODE 5 and MODE 6 (limited) to meet the FSAR requirements in reference 8, 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 talve is resolved by the expected Operability of the RWST providing 5 hours of injection flow. The recovery description also means that the RWST must be available in MODE 6 until the vessel head is removed i and the refueling canal is filled to its specified 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 MODE 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: With the RWST Inoperable, restore the tank to OPERABLE status within 1 hour, or be in at least HOT STANDBY [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 hours and in COLD SHUTDOWN within the following 30 hours. The Licensing Basis FSAR requires Safety Injection of 2000 ppm Boron to mitigate the nuclear power consequences of any accidents which may initiate during this , period; if the RWST is not available, then Boron Concentration in the RCS should I be increased to the level required to mitigate any potential return of nuclear power. The proposed T.S. appears nonconservative. The licensee shall evaluate and propose and in so doing he should evaluate each , of the Operability requirements separately to determine if COLD SHUTOOWN is l required for each INOPERABILITY REQUIREMENT, or whether alternate mitigating l Actions are possible. l 06/01/84 95 Revision A

 \                                                                                                      ...
   ~

T.S. Section 3/4.7 PLANT SYSTEMS - T.S. Page 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 generator shall be OPERABLE with lift settings as specfied in Table 3.7-3. MODES 1, 2, and 3. APPLICABILITY: 1 ACTION: l

i

a. With four reactor coolant loops and associated steam generators in operation and with one or more main steam line code safety valves inoperable, operation in MODES 1, 2, and-3 may proceed provided, that within 4 hours, either the inoperable valve is restored to OPERABLE status or the Power Range Neutron Flux High Trip Setpoint is reduced per Table 3.7-1; otherwise, be in at least HOT STANDBY within the next

6 hours and in COLD SHUTDOWN within the following 30 hours.

b. With three reactor coolant loops and associated steam generators in operation and with one or mora. main steam line code safety valves associated with an operating loop inoperable, operation in MODES 1, l 2, and 3 may proceed provided, that within 4 hours, either the inoperable valve is restored to OPERABLE status or the Power Range i Neutron Flux High Trip Setpoint is reduced per Table 3.7-2; otherwise, be in at least HOT STANDBY within the next 6 hours and in COLD SHUTDOWN with the following 30 hours.

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 i for Reactor Protection through its impact on SG/RCS system response under l Condition II, III, and IV occurrences. It also affects the integrity of the Primary Containment Boundary. We do r.ot find an adequate consideration of the alternate type of Safety Valve Failure that can occur, and their related significance, upon the action state-

     .              ments proposed.                                                                           -

How sure is the Licensee that inadequacy to meet the very limited single operability 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 the severe conditions of Transient and Accident Events. We find the proposed T.S. inadequate in its representation of operability, or lack there of, for these Safety Valves. Consequently, without a requirement that they all be operable in MODES 1, 2, 3, and 4, with a further requirement l l l 06/01/84 96 Revision A ; 1 1 I L______.____________

V ____-__ - _ - - W 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. Page 3/4 7-4: AUXILIARY FEEDWATER SYSTEMS Item: APPLICABILITY MODES 1, 2 and 3 in the proposed T.S. should be expanded to MODES TATION,4Itemsand 57ina,accordance with b, c, d, e, and f. our review under Table 3.3-3 ESFAS INSTRUMEN-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. Item 3.7.1.2.b. The licensee has deleted OPERABILITY requirements for the Staam-Turbine driven auxiliary feedwater pump at steam pressures of less than 900 psig. This provided: cation has been is not in accord with current Accident Analyses and no justifi-Reference 15, Recommendation GL-3, requires the Steam-Turbine AFW pump in the event of complete loss of AC power for a period of 2 hrs and beyond. This will require operability down to the lowest pres-surer for which the Turbine is provided as described in reference 22, Table 10.4.7-6 where the range of operating pressures provided for is from 110 psig to 1205 psig. This will also provide for operabilty down to and including MODES 4 (and availabiilty f rom MODE 5) to cover licensing require-nents discussed elsewhere under Table 3.3-3, ESFAS INSTRUMENTATION, Items 7a through f. We note two principal features relating to the service conditions of the Turbine 9 riven Feedwater Pumps: a. They are supplied with steam from two steam generators from main steamGenerators, Steam lines af ter the flow restriction orifices at outlets from the b. They would normally be expected to perform early in the transient and continue to function to design flow requirements throughout the Occurrence. The licensee should explain how the proposed TS ensures that the Turbine Driven pump maintains its flow performance required by Accident Analysis when steam line pressures could drop substantially 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 drift and errors). The licensee shall evaluate the above comments and propose technical specifi-cations which will ensure operability of the Turbine-Oriven AFW Pump over the range of conditions expected from Design Basis Accident Analysis, and other less bounding events, down to and including MODE 4 as discussed in the Licensing Basis. In his evaluation, the licensee should advise if Item le of Table 3.3-5 ESFAS INSTRUMENTATION, Steam Line-Pressure Low is derived from steam line sensors and after the SG orifices, or if it is taken from pressure sensors on the Steam Generator. The licensee should then advise what has been used in assessing Steam Generator Pressure Response and Turbine Driven AFW pump response in the 06/01/84 97 Revision A

                                                                                                                .      l Condition III and especially Condition IV Occurrences of the. Licensing Basis, and if the existing Accident Analyses remain valid.                                                           ,

Jtem 4.7.1.2: SURVEILLANCE REQUIREMENTS The Technical Specifications, page T.S. 3/4 7-4 requires each motor driven (MO) AFW pump to supply 450 pgrr. 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 Accident Analysis on j page 15.4 - 13 a (Revision 38) in which the MDAFW pump headered to two intact ] steam generators supplies 170 gpm each whilst the-one headered to the faulted ] Steam Generator suppies 110 gpm to the 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 both these SGs supply the Turbine Driven AFW pump. The Licensee committed to establish and verify by test, the valve throttle positions neces-sary to achieve 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 that prior to plant startup following an extended shudown, a flow test would be performed to verify the normal flowpath from the primary AFW , system to the steam generator. The flow test should be conducted with AFW l system valves in their normal alignment. At this time, we do not see a proposed T.S. which ensures that the required subdivision of flow between 3 intact and 1 faulted steam generator, and 2 Intact and 2 " Faulted" Steam Generators associated with the Turbine-Oriven AFW Pump, required by the Licensing Basis is achieved, and we do not see any test period recommended such as following an extended cold shutdown to ensure that the required flow division is maintained in an acceptable manner. At this time we must conclude 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-Sc Proposed: CONDENSATE STORAGE TANK SYSTEMS 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 shall be operable with a contained water volume of at least 175,000 gallons of water. 06/01/84 98 Revision A

5 APPLICABILITY MODES proposed are 1, 2 and 3, with lesser volumes required in , , MODES 4 and 5. ACTION STATEMENT should include a provision that, with the condensate's'torage

       . tank inoperable, within 4 hours either                                                        '

a. Restore the CST to OPERABLE status or be in at least HOT STANDBY I within the next 6 hours and in HOT SHUTDOWN within the following 6 hours, 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 pumps, and restore the condensate storage tank to OPERABLE status within 7 days, or be in at least HOT STANDBY within the next 6 hours and in HOT SHUTDOWN within the following 6 hours.

SURVEILLANCE REQUIREMENTS should include i

a. The condensate storage tank system shall be demonstrated 0PERABLE at least once per 12 hours by appropriate measures when the tank is i the supply source for the auxiliary feedwater pumps.

b. The Nuclear Service Water System and Standby Nuclear Source Water Pond shall be demonstrated OPERABLE at least once per'12 hours by appropriate measures l Additionally, an evaluation of and provision will need to be made concerning potent,ial loss of AFW supplies during loss of suction and change-over to alternate AFW sources.

The safety basis for these requirements are

a. Our earlier review under TS. Table 3.3-5 Items 7a and 7b show that whereas all safety evaluations involving AFW supply have assumed a Safety Analysis Limit of 61 sec. response time, this is only available from nonsafety related water sources. Further, that the safety l

related supply from the Nuclear Service Water Pond may take an extra

15 secs which is substantially non-conservative in respect of the related safety analysis. Therefore, at this time, until the licensee has evaluated our concerns and made - acceptable proposals, the NRC will require technical specifications on this non safety-related water storage of the above nature. The proposed T.S. are nonconservative with respect to Regulatory Requirements. The licensee shall evaluate and propose. l T.S. Page 3/4 7-8: MAIN STEAM ISOLATION VALVES Item 3.7.1.4 The proposed T.S. provides that: "each main steam line isolation valve (MSLIV) shall be OPERABLE with APPLICABILITY MODES 1, 2, and 3. 06/01/84 99 Revision A

v 4 The requirements within the Licensing Basis for Main Steam Line Isolation are (q discussed in this review under Table 3.3-4, Item 4. The Licesing Basis does require operability in 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" gg and the Licensing Basis FSAR under referencef4 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.
                                                            ~~

T.S. Page 3/4 7-8a Proposed: STEAM GENERATOR POWER OPERATED RELIEF VALVES (SG PORVs) The proposed T.S. does not include these valves which are required to enable l 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 , of 10% Rated, and the review under Table 3.3-1 under Item: Concerning Prescribed l - Values for % Rated Thermal Power DURING ST. ART UP (MODE 1)_ AND POWER OPERATION (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 available to meet this situation. 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 available l 45% (of full load main steam flow to atmosphere) described under reference 22, I page 10.1-2, revision 8, para 3, if they can be available under loss of Offsite Power. An appropriate Action Statement should be provided. If the additional 64 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

to the SG PORVs is contrary to Regulatory Requirements which have been excluded CT from the Licensing Basis. The Licensee shall evaluate and propose. -" T.S. Section 3/4.7.3: COMPONENT COOLING WATER $YSTEM The proposed T.S. requires that: 3.7.3 At least two independent component cooling water 1 cops 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 hours or be in at least HOT STAND 8Y within the next 6 hours and in COLD SHUTD0km within the following 30 hours. The SER for the plant under reference 10, summarizes the following Licensing Basis for the Component Cooling System: 9.2.4 Component Cooling System The component cooling system provides cooling water to selected nuclear auxiliary components during ,ormal 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 reactir 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 poo! water; and (4) pr;v1Je cooling to dissipate waste heat from various

primary station components during normal operation and postulated accident conditions. Active system components necessary for safe plant shutdown are designed to include at least 100 percent redundancy. The component cooling water for each unit includes two component cooling heat exchangers, four component cooling pumps and a split-volume component cooling surge tank. Two pumps and one heat exchanger per unit provide the necessary cooling water for normal operation, cooldown, refueling, and postulated accidents. The remaining pumps and heat exchangers serve as standby. An assured supply of makeup is provided from the nuclear service water system to each redundant loop. The component cooling water system is designed to seismic Category I requirements, except for certain branches to non-essential equipmont. - The component cooling water pumps are powered by redundant emergency buses. The portion of the component cooling water system serving the residual heat removal system meets the single failure criterion for active components. Based on our review, we conclude that the component cooling system design is in conformance with the requirements of General Design Criterion 44 06/01/84 101 Revision A

of Appendix A to 10'CFR Part-50 regarding the capability of the system to ) transfer heat from systems and components important to safety to an ' i ultimate heat sink and provisions of suitable redundancy for safe cool- ' down. We further conclude that.the system design meets the requirements of General Design Criteria 45 and 46 of Appendix A to 10 CFR Part 50 ' regarding system design that allows performance of periodic inspections l and testing. We conclude that the component cooling. water system is l acceptable.

                -                                                                                                    1 Detailed reference to Operability and Operating requirements in the Licensing Basis in MODES 5 and 6 can be found in reference'22, page 92-17 and Component-Cooling System.

l The proposed T.S. completely ignores, without any evaluation, the Licensing j fp Basis requirement for this system in MODES 5 & 6. The current T.S. are non ' ! conservative with respect to the Licensing Basis. The Licensee shall evaluate and propose. This T.S. is a prime example of a Standard Technical Specification which completely ignores the Licensing Basis for all Nuclear Power Plants. This reflects a very serious Safety Issue for all standard T.S. and which cannot await an extended " Generic" Resolution. T.S. Section 3/4.7.4 NUCLEAR SERVICE WATER SYSTEM APPLICABILITY MODES proposed are 1, 2, 3, 4. These should be extended =to (f 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 shutdown." The equipment requiring NSWS also includes all RPS and ESFS systems, many of which are necessary in MODES 5 and 6 to the above redun-dancy and single failure criteria. Examples include: MODE 5 is required to service AFW alternate cooling require-ments in event of a fail-closed RHR/RCS isolation valve in the RHR line, and in MODES 5 and 6 it is needed to service necessary redundant RHR Trains. Reference our related evaluations in this review concerning RHR operability requirements in MODES 5 and 6. The proposed T.S. is nonconservative with respect to the Licensing Basis. .The - I l licensee shall evaluate and propose. T.S. Section 3/4.7.5 STANOBY NUCLEAR SERVICE WATER POND (SNSWP) , 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 than 94 F l

                                 ....in the intake structure"                                                        ,
 .                                                                                            .                   j 1

4 4 l 06/01/84 102 Revision A i l I

W , 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 Components including NPSH requirements (reference 6, page 9.2-13, last para). An average water temperature of 70 F has been selected by RSB as a potential design basis for Condition II, III and IV occurrences. The licensee has pro- j vided little information on the range of AFW temperatures used in his analyses i and the related sensitivity of results to AFW 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 temperature 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 temper-atures for water supply. 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 temperatures 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 temperatures in the standby 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 hear-freezing temperatures must be considered as nonconservative in respect of certain events. The ' Licensee shall evaluate and propose. J APPLICABLE MODES: The system is required in all MODES 1, 2, 3, 4, 5, & 6 to b handle heat-rejection requirements as the ultimate heat sink. The licensees proposal to limit this to MODES 1, 2, 3 and 4, is nonconservative with respect to the Licensing Basis. The licensee shall evaluate a.,nd propose. - Reference 6, page 9.2-13, revision 39, states that "In the event of solid ! layer of ice" forms on the SNSWP, the operating train [of the Nuclear Service Water [NSW) system] is manually aligned to the $NSWP. The Licensee shall provide the Safety Related reason for this action and advise if this operator action conflicts with the Response Times proposed.under Table 3.3-5. Given a Safety Related reason, surveillance requirements ensuring this action should be included under either T.S. Section 3/4.7.5 NSWS or this particular T.S. Section 3/4.7.5 STANDBY NSWP. Absent this, surveillance requirement on a , Safety Related Issue, the proposed T.S. would be non-conservative. The Licensee shall evaluate and propose. 06/01/84 103 Revision A \ _ ____ ____-___ ___ _______ _ :

v I T.S. Section 3/4.9 REFUELING OPERATIONS l l T.S. Item 3/4 9.1 BORON CONCENTRATION l l 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 boron dilution analyses of this reference 7, provides that, during refueling: j

a. "A minimum water volume in the Reactor Coolant System is considered.

This corresponds to the volume necessary to fill the reactor vessel ' above the nozzles to ensure mixing via the residual heat removal J l 1oop." l l b. Neutron sources are installed in the core and the source range I detectors outside the reactor vessel are active and provide an audible count rate. l c. A high flow alarm at the discharge of the CVCS (from flow element ! INVFE 5630) is active providing an alarm to the operator when the l flow rate from the charging pumps exceeds 175 gpm.

d. The charging pumps are inoperative.

Additionally, 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 or above l j the loop provide acceptable flow, . including NPSH conditions, at inlet to the l RHR pumps, j 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. i The current SER, Supplement No.1, reference 11,15-1, provides that:

                                                                   "During refueling the applicant has committed to isolate all sources of                       1 unborated water connected to the primary system refueling / canal / spent fuel.

We do note that Surveillance Requirement T.S. 4.9.1.3 does provide for verifying that valve No. INV-250 is closed, under administrative control in support 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 t                                                                   the volume control tank with the other bypassing this tank. These

lines contain flow control valves INV171A and INV175A respectively."

l 06/01/84 104 Revision A 1

~ l- , Why are T.S.s not applied to the closure of these valves also. The proposed T.S. may be nonconservative with respect to the Licensing Basis. The licensee shall evaluate and propose. We also note an apparent non-conservative discrepancy between the basis for the specified reactivity condition of "a k of 0.95 or less" without-any specificatiunofthepositionofmovablecbrolassemblies. We also note the-need to add, according to reference 7, page 15.2-14, revision 10, that the boron 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 requiremer t underlined should be a part of the LC0 for this T.S. item. Without this pro- < vision in the proposed T.S, it could be interpreted as non-conservative in respect of the Safety Analysis Limits for the plant. The license., shall evaluate and propose. In the Licensing Basis FSAR reference 8, page Q 212-24, item 212.57, it is required that the reactor makeup water pumps shall be removed from the loads supplied by the emergency power supplies. This is to prevent inadvertent boron - dilution during certain Occurrences in which electrical loads are disconnected from, and returned to, the Emergency Buses. Provision should be made 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 shall evaluate and propose. T.S. Item 3/4 9.8 RESIDUAL HEAT REMOVAL AND COOLANT CIRCULATION; HIGH WATER LEVEL The LCO provides that: 3.9.8.1 At least one residual heat removal (RHR) loop shall be OPERABLE and in operation.* The Licensing Basis, reference 20, Page 5.5-23, under Refueling, and page 5.5-24 under 5.5.7.3.1, System Availability and Reliability, last paragraph, shows the licensing of the RHR system is never based on only one RHR system being operable. Two are always to be availabl E This proposal is therefore outside the LC0 for the FSAR in a non-conservative manner. The Licensee shall evaluate and propose In his Basis, on T.S. Page 3/4 9-2, last para. , the licensee has proposed that:

 .                                                                                                                  ~
                                     "With the reactor vessel head removed and 23 feet of water above the reactor vessel flange, a large heat sink is available for core cooling.

Thus, in the event of a failure of the operating RHR loop, adequate time is provided to initiate emergency procedures to cool 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 number of alternates are proposed to resolve the situation and almost l invariably, electric power is required, and in most cases the RHR equipment is I used. If the basis for the licensee's request here is to enable him to operate

                           .06/01/84                                   105                               Revision A

m 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 c isolation required in 2 hours, has not been evaluated. At this time we have no acceptable safety basis for allowing the proposed deviation from the Limiting Conditions of Operation of the Licensing Basis FSAR which is that 2 RHR loops . from 3eparate emergency buses be operable. The proposal is therefore non-conservative and the licensee must evaluate and propose. Furthermore, the licensee must provide that the level of water in or above the loops be such as to provide acceptable flow, including NPSH conditions, at { inlet.to the RHR pumps. Absent those required conditions from the Limiting l Conditions of Operation could make them non-conservative. The licensee shall evaluate and propose. Ch The ACTION STATEMENT provides that with no RHR loop operable, the containment g should be closed within 4 hours. Information in reference 8, page Q 212-56 under Case'2 shows that if RnR is absent [by isolation of the RCS/RHR inlet valve] that: l "Approximately 2.5 hours are available to the operator to establish an-alternate means of core cooling. This is the time it would take to heat 300,000 gallons of water in the refueling canal from 140 F to 212 F,. ; assuming the maximum 24 hours. decay heat load." The current value of 4 hours appears less conservative tnan this calculated value of 2 hours within the FSAR. The licensee shall evaluate and propose. The current surveillance requirement: 4.9.8.1 "At least one RHR loop shall be verified to be in operation and I circulating reactor coolant at a flow rat, of greater than or equal to

                 .                  3000 gpm at least once per 12 hours."

is deficient in that the thermal performance of any one RHR system to Licensing Basis safety requirements is not being verified. The T.S. is therefore non-conservative with respect to the Licensing Basis. The licensee shall evaluate and propose. --- Footnote *: The licensee also proposes that, "The (only operable] RHR loop may be removed from operation for up to 1 hour per 8-hour period during the performance of CORE ALTERATIONS 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 and in so doing decrease the available response to handle a loss of cooling capacity from 2 hours down to 1 hours, and for a considerable period of time

thereafter whilst temperatures aro again being reduced to the required value of 140 F.] This proposed T.S. is outside the Licensing Basis in a nonconserva-tive manner. The Licensee shall evaluate and propose. 06/01/84 106 Revision A _ _ _ _ _ _ _ _ - _ _ _ _ .o

m .- . 4 e Review of available responses to the consequences of a fail closed RCR/RHR G

                                                                                                              .y 1

isolation valve, include any procedures using the containment sump. To allow W1 for this single failure contingency, the licensee should therefore ensure that the containment sump will be operable during this mode, and with an appropriate 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. The licensee shall evaluate and propose. T/S Page 3/4 9-12 REFUELING OPERATIONS l 1 The subtitle should read as 3/4.9.9 HIGH WATER LEVEL Clarify by addition of the term HIGH I T/S Page 3/4 9-11 REFUELING OPERATIONS LOW WATER LEVEL 1 APPLICABILITY: MODE 6 when the water level above the top of the reactor vessel flange is less than 23 feet. l GENERAL REVIEW: Whereas the existing FSAR under reference 20, page 5.1-7 ; discusses Refueling, it does not provide for a sustained period of normal . operations under these Low Water Level conditions. The FSAR provides that: l

                  " Refueling                                                                                           i l

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 below the reactor vessel flange. The vessel head is then raised as the refueling canal is flooded. Upon completion of refueling, the system is refilled for startup." Furthermore, we find that the FSAR analyses of the single failure of the l 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 developed q l for these circumstances. However analyses have been undertaken for the water ; inventories and temperatures in the RCS system that might apply under those i conditions. Presumably therefore, the "0PERATING MODE - LOW LEVEL" is a long l term changing condition following Cold Shutdown, with loops drained and bolts , tensioned changing to bolts untensioned and removal of the head, as concomitant _ l flooding of the reactor vessel cavity continue:. 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. I Further, since this is effectively a long term changing condition, in the FSAR, . 'g .: ." it is not acceptable to allow some of the provisions requested g such as one '

i. j ;..

hour for the performance of CORE ALTERATIONS--which b) T.S 3/4 9.9-fare q, only 'g . ".' permissible under that specification with at least'23 feet of water over the reactor vessel flange. V

9 l l e%

                                                                                         -'    -% ~!        .

l 06/01/84 107 Revision A

e M Q. ' It.is proposed that an additional. item be added.to the current statement of This MODE shall not to be used for continuous APPLICABILITY to the effect that: - normal operations, but only as a set of circumstances occurring during the 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 are being drained for replacement and tensioning of the Reactor Vessel Head. The licensee shall evaluate and propose. The existing LCO specifies that: 4 "3.9.8.2 Two independent residual heat remval (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

        . power.from a different source; reference 20, page 5.5-24, revision 9. Without this requirement, the T.S is less conservative than the FSAR and the licensee shall evaluate and propose.

p Additionally, the current FSAR, reference 8, page Q212-57, revision ~25, describes 9 that'in the event of loss of flow caused by closure of the RHR/RCS isolation ' i 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 flow alarm.                           -

Assuming worst case conditions (maximum 24 hours 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 off,

                .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 loop is required to be in operation, the LCO should ; therefore require 2 operable safety related RHR low flow alarms on each single j operating system so that the operator can respond within 10 minutes to commence i operation of the redundant system. Is this time frame excessive since boiling will have commenced. It is necessary to maintain two operating.RHR systems so that boiling will not occur with a single failure. The licensee shall evaluate

      , and propose.

Additionally, the above information defines an LCO of a minimum volume of water - for the related event in which the RCS is drained to just below the level flange. A further requirement (LCO) is that any such minimum volume should be such that the level of water in or above the loop provides acceptable flow, including NPSH conditions, over the range of temperatures expected at inlet to the RHR pumps. Absent those required conditions from the Limiting Conditions of Opera-tion makes them non-conservative in respect of the Licensing Basis. The licensee shall evaluate and propose. 06/01/84 108 Revision A

Footnote *: provides that, (R5N

                            "* Prior to initial criticality the RHR loop may be removed from opera-tion for up to 1 hour per 8-hour period during the performance of CORE 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 non-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 par 12 hours."

l A time delay of 12 hours is excessive to verify a loop in operation, and this has been considered earlier in this section. Further, the surveillance requirement,- every 12 hours, is intended to ensure not only that the system is operating, but,that it is operating at p'rocess 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 requirements for this item are absent most of this information; it is therefore non-conservative and the licensee shall evaluate and propose.

dr The current ACTION STATEMENT calls for containment closure in 4 hours [i.e. 240 mins]. Earlier conservative calculations for this MODE show that loss of g all RHR in this MODE can cause boiling in 5 minutes and core uncovery in 100 mins. Given the circumstances, containment enclosure should be effected immediately, commencing RHR low flow alarms. The licensee shall evaluate, and propose. The current T.S. appears nonconservative with respect to the Licensing Basis. - l I l 1 . 06/01/84 109 Revision A t

O Addenda . T.S. SECTION 3/4.5 EMERGENCY CORE COOLING SY' STEMS T.S. SECTION 3/4.4.4.1 RCS LOOPS AND COOLANT CIRCULATION / HOT SHUT 00WN MODE 4 gp More recent information, and a detailed check on certain elements of the proposed T.S. relevant to the above section, and the Licensing Basis FSAR, ( 83dI) and particularly reference 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 l acceptable surety that: ) a) The Reactor Coolant Pressure Boundary (RCPB) valves on the RHR/RCS suction line are confirmed closed in MODES 1, 2, & 3. b) That the RCPB valves in the RHR/RCS suction line are individually identified as opened in the RHR MODE. c) That in RHR MODE 4,the RHR system must be capable of automatic re-alignment to the ECCS mode with residual ECCS equipment, in the event of a SI signal, including automatic closure of the RCPB Isola-tion valves on the RHR/RCS Suction Line in accordance with 10 CFR 50 App A Criterion 55(4) and subsequent automatic opening of valves to the RWST in accordance with 10 CFR 50 App A, Criterion 20 [with appro- l priate provision for RHR pump protection]. The current position in respect of c above appears to be absent those-

                                                         . requirements and therefore non-conservative. The Licensee shall evaluate and propose.                    '

The T.S. should provide the LCOs and surveillance in the overpressurization protection system of the RHR system as described in Licensing Basis FSAR, reference 3, page 5-5-24. l l Proposed T/S P' age 3/4 5-6, item 4.5.2.d, 1) b) appears incorrect: it provides l that, in establishing ECCS operability: l

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 Syste:n pressure signal greater than or equal to 425 psig the interlocks prevent the valves from being opened, 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) above is incorrect in that it should ensure that with a simulated or actual Reactor Coolant System pressure signal greater than 475 psig, the 06/01/84 110 Revision A

t interlocks will cause the valves to automatically close, reference 4 section 5.5.7.3.3 and reference 5, section 7.4.1.5.4. o.grir %eM am,,yM

 //gp he proposed T.S. closes the valves when they are in fact required to be ooen and is>therefore non-conservative. Further, the lower pressure of p/h ] 475 psig required to close is more conservative than a valve of 560 unless Inere are Set Point and Channel considerations - The pressure is less conser-vative than the Licensing Basis F5AR value.

l

                                                                                                            .- s ,.

I 06/01/84 111 Revi sion JC/f,

4 interlocks Sill cause the valves to automatically close, reference 4, l section 5.5.7.3.3 and reference 5, section 7.4.1.5.4. The proposed T.S. closes the valves when they are in fact required to be

     .           open and is therefore non-conservative. Further, the lower pressure of 475 psig required to'close is more conservative than a valve of 560 unless there are Set Point and Channel considerations - The pressure is less conser-                                                                                                                 ,

vative than the Licensing Basis FSAR value. -f 1 a

J l I l j l . - l 06/01/84 111 Revision A _.____ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ . _ _ . _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ . . . _ . _ _ _ _ _ . . __ . _ _ . . _ _ . . . - _ .__._____.-______________1_.

v d' l . LIST OF REFERENCE _5

1. Letter f rom H. B. Tucker (D.P.Co) to H. R. Denton (NRC) dated September 27, 1 1982 to the subject of "McGuire Nuclear Station."

2. Memo from C. O. Thomas (SSP 8) to Brian W. Sheron (RSB) on the subject of i

           " Proof and Review of McGuire - Units 1 and 2, Technical Specifications."                  {

Dated January 14, 1983.

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, I Duke Power Company, McGuire Nuclear Station, Units 1 and.2, Rev. 45. 5

5. U.S. Nuclear Regulatory Commission, Final Safety Analysis Report, Volume 7, , l Duke Power Company, McGuire Nuclear Station, Units 1 and 2, Rev. 45. !

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, l 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 Docket 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 Docket Nos. 50-369 and 50-370,-

May 1978.

12. U.S. Nuclear Regulatory Commission, Office of Nuclear Reactor Regulation, I

           " Safety Evaluation Report, McGuire Nuclear Station, Units 1 and 2, Duke
.          Power Company," NUREG-0422, Supp. No. 2, on Docket 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 Company," NUREG-0422, Supp. No. 3, on Docket 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 Docket Nos. 50-369 and 50-370, January 1981.

06/01/84 112 Revision A

m Y

      *'   {,

j- 4

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 5 and 2". Dated February 08, 1983.

17. Letter from H. B. Tucker (D.P.Co) to H.~R. Denton (NRC) on the subject of McGuire Nuclear Station, Units l'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 (D.P.Co) to H. R. Denion (NRC), dated Oct. 8, 1981 on the subject of McGuire Nuclear 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 dated April 1981, published by Westinghouse Electric, Nuclear Energy Systems, ; PROPRIETARY. '

19. Westinghouse Electric Corporation, PWR Systems Division " Westinghouse i Emergency Core Cooling System - Plant sensitivity studies, WCAP-8356. '!

August 1,1974. ^ i

20. U.S. Nuclear Regulatory Commission, Final Safety Analysis. Report, Volume 4, 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 Anal'ysis Report, Volume 9, .I 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), dated August 13, 1980, re: McGuire Nuclear Station.

24. Letter from W. O. Parker (0.P.Co) to H. R. Denton (NRC), dated September 18, 1980, re: McGuire Nuclear Station. Page 13, Response to 3(e).

25. Duke Power Company McGuire Nuclear Station, Unit 1, Docket No. 50-369, License No. NPF-9 Startup Report, February 15, 1982.

I

26. Memo for RSB, CPB, ICSB Members from Brian W. Sheron (RSB), Carl H. '

8erlinger (CPB), Faust Ross (ICSB) dated April 12, 1983 on the Subject of Inadvertent Boron Oilution Events.

27. Westinghouse Electric Corporation, Nuclear Energy Systems Topical. Report, Overpressure Protectior for Westinghouse Pressurized Water Reactors, WCAP-7769, Rev. 1, June 1972. ,

j 06/01/84 113 Revision A J l

Q

28. Westinghouse Electric Corporation for the Westinghouse'0wners Group on Reactor Coolant' System Overpressurization, July 1977. -

29. U.S. Nuclear Regulatory Commission, Final Safety Analysis Report, Volume 6, Duke Power Company, McGuire Nuclear Station, Units 1 and 2, Rev. 45.

s I i 06/01/84 114 Revision A

w

                                                    ,i                                                  ,j
                                                                                                        ,                         1 q
                                                                                                                                                                       .                                  )'

TABLE 1-SECTIONS REVIEWED BY REACTOR SYSTEMS BRANCH

                                             -SECTIDN                                                                                                                           ,

PAGE 1 2.1 SAFETY LIMITS i-2.1.l aREACTOR CORE ................................................... 2-1 2.1.2 REACTOR COOLANT SYSTEM PRESSURE ................................ 2-1 ! FIGURE 2.1 L1 REACTOR. CORE SAFETY LIMIT - FOUR LOOPS IN OPERATION . . . . . 2-2 2.2 LIMITING SAFETY SYSTEM SETTINGS , ,

                                                                                                                                                                                                                     .q 2.2.1     REACTOR TRIP SYSTEM INSTRUMENTATION SETP01NTS . . . . . . . . . . . . . . . . . . .                                                          2-4              1 l

TABLE 2.2-1 REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS ..... . . 2 3 /4. 0 A P P L I C A B I L ITY . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 /4 0-1 3.4.1 REACTIVITY CONTROL SYSTEMS 3/4.1.1 B0 RATION CONTROL . i Shutdown Margin - T 3/4 1-1 avg > Programmed No load Tavg ........... ]; Shutdown Margin - T < Programmed No load T 4 and > 200 F . . avg avg. 1

Shutdown Margin - T < 200 F .. . .......... ... ...... . 3/4 1-3 avg Moderate Temperature Coefficient ................ ..... .... 3/4 1-4 1 Minimum Temperature for Criticality .........

                                                                                                                                                                    ....      ........          3/4 1-6 l

3/4.1.2 BORATION SYSTEMS- l Flow Path - Standbye, Shutdown and Refueling ......... . .... 3/4 1-7 Flow Paths - Power Operation, Startup, Standbye down to 1000 psig/425 F .................................. ... 3/4 1-8 -

       '                                                                                                                                                                                                               i 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 t Borated Water Sources - Operating ............... ... . , . 3/4 1-12 ' Instrumentation ............. . . . ...................... 3/4'l-13e 05/01/84 115 Revision A l E_ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _

       *t                               .
                                                                                                                                 ~

Y. < SECTION-PAGE l l TABLE 3.1-l' ACCIDENT ANALYSES REQUIRING REEVALUATION IN THE EVENT { OF AN INOPERABLE FULL-LENGTH RUD ..'...................

                                                                                     .                                3/4 1-16                   I l

3/4 1-17 Position Indication Systems - Operating ......... ..........

                        . Posi tion Indication System - Shutdown . . . . . . . . . . . . . . . . . . . . . .          3/4 1-18 Rod Orop Time (Units 1 and 2) ......................... ....                                 3/4 1-19                 3 l                         Shutdown Rod Insertion Limi t (MODES 1 & 2)' . . . . . . . . . . . . . . . . .               3/4 1-20 l

Shudown Rod Insertion Limits (Modes 3 - 5) ................. ) 1 i Control Rod Insertion Limits ... .. ................... .... 3/4 1-21 > ' a

           . 3/4.2' POWER DISTRIBUTION LIMITS TABLE 3.2-1      DNB AND REACTOR COOLANT SYSTEM PRESSURE PARAMETERS . . . . . . -3/4 2-16 I

3/4.3 INSTRUMENTATION -) i 3/4.3.1 REACTOR TRIP SYSTEM INSTRUMENTATION ............. . . . . ' . . . . . . . 3/4 3-1 l TABLE 3.3-1 REACTOR TRIP SYSTEM INSTRUMENTATION ............. ... ... 3/4 3-2 TABLE 3.3-2 REACTOR TRIP SYSTEM INSTRUMENTATION RESPONSE-TIMES .... . 3/4 3-9 i TABLE 4.3-1 REACTOR TRIP SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS .. .. ............ ...... ............. 3/4 3-11 3/4.3.2 ENGINEERING SAFETY FEATURES ACTUATION SYSTEM l INSTRUMENTATION . .... ........................ .. ..... 3/4 3-15 1 , TABLE 3.3-3 ENGINEERED SAFETY FEATURES ACTUATION SYSTEM

                                ' INSTRUMENTATION .......r7:...........................                            . 3/4 3-16                   ,

TABLE 3.3-4 ENGINEERED SAFETY FEATURES ACTUATION SYSTEM , INSTRUMENTATION TRIP SETPOINTS .................. ..... -3/4 3-25

    .        TABLE 3.3-5      ENGINEERED SAFETY FEATURES RESPONSE TIMES ...............                               3/4 3-30              _

4 3/4.4 REACTOR COOLANT SYSTEM 3.4.4.1 REACTOR COOLANT LOOPS AND COOLANT CIRCULATION l Startup and Power Operation ... ............................ 3/4 4-1 1 Hot Standby ..... .. .............................. ........ '3/4 4-2 i Hot Shutdown ..... ... .............. ...................... 3/4 4-3 1 06/01/84 116 Revision A j 1

     ~ - -

l SECTION 1 PAGE Cold Shutdown '- Loops Filled .. . . . . . . . . . .... ............ -3/4 4-5 Col d Shutdown - Loops Not Filled . . . . ' . . . . . . . . . . . . . . . . . . . . . . - 3/4 '4-6 3/4.4.2 SAFETY VALVES Shutdown ................................................... 3/4 4 Opetating .................................................. 3/4 4-8 i 3/4.4.3 PRESSURIZER ................................................ 3/4 4-9 3/4.4.4 RELIEF VALVES ............................................. 3/4 4-10 3.4.4.5 STEAM GENERATORS ........................................... 3/4 4-11. !

.]

3/4 4-35 Pressurizer ................................................ Overpressure Protection Systems ............................ 3/4 4 3/4. 5 EMERGENCY CORE COOLING SYSTEMS-3/4.5.1 ACCUMULATORS 1

            .                                   Cold Leg Injection ............................./...........                                         3/4 '5-1 Upper Head Injection         ..........               .. ......... ........ ....                     3/4 5-3    .

3/4.5.2 ECCS SUBSYSTEM - T 3/4 5-5 avg > 350 F ... ... .. ................ 3/4.5.3 ECCS SUBSYSTEMS - T < 350 F ... ........... ............. 3/4 5-9 avg 3/4.5.4 BORON INJECTION 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 Turbine Trip on-Reactor Trip'......... . .. .. 3/4 7-1 ; Auxiliary Feedwater System .... ............................ 3/4 7-4 Auxiliary Feedwater Condensate Storage System ........... . 3/4 7-5(a) l l Main Steam Line Isolation Valves . .......................... 3/4 7-8 , Atmospheric Oump Valve ....r.. ..... ... ............... . 3/4 7-8a 3/4.7.2 STEAM GENATOR PRESSURE / TEMPERATURE LIMITATION .............. 3/4 7-9 06/01/84 117 Revision A

e SECTION PAGE 3/4.7.3 COMPONENT COOLING WATER SYSTEM . . .... .......... . 3/4 7-10 3/4.7.4 NUCLEAR SERVICE WATER SYSTEM ..... .................... ... 3/4 7-11 - 3/4.7.5 STANDBY NUCLEAR SERVICE WATER POND .......... .. .. .... . 3/4 7-12 3/4.9 REFUELING OPERATIONS 3/4.9.1 BORON CONCENTRATION ........... ... ... .......... ........ 3/4 9-1 3.4.9.2 INSTRUMENTATION ..... ........................ ......... .. 3/4 9-2 3 3/4.9.8 RESIDUAL HEAT REMOVAL AND COOLANT CIRCULATION f 1 High Water Level ........ ................. ....... ..... 3/4 9-10 Low Water Level ...... .. .... .... .............. ... . ... 3/4 9-11 l l [ i 1

_4 i

t l 1 f 06/01/84 118 Revision A _____-_L_ __ - - ..

TABLE 2 l TECHNICAL SPECIFICATION PAGES AFFECTED The following pages of the Technical Specifications are affected by this review: T.S. Pages 2-1, 1 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 i l 3/4 1-7 I 3/4 1-8 3/41-9 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 i 3-4 l 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 l 3-20 l 3-21 3-22 3-23 ' 06/01/84 119 Revision A

   -:                                                                                                                             r--

4 f TABLE 3.3-4, T.S. Pages , 3/4 3-25 , 3 3-27

                                                                                   .3-28 3-29                                                    0 TABLE 3.3-5, T.S. Pages                                   3/4 3-30'
                                                                                                                                            )
                                                                                                                                            )

3-31 3-32 3-33 T S', Pages 3/4 4-1 4-2 , 4-3 4-4 4-5 1 4-6 4-6(a) proposed 4-7 4-8 4-9 4-10 4-11 4-36

            'T.S. Pa'ges                                              3/4~5-1, 5-2 5-2a) proposed 5-2b)' proposed 5-3 5-4                                                       l 5-4a) proposed-                                         .]

5-4b) proposed l 5-5' l 5-6. l 5-8 5-9 5-10 5-11 5-12 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 o T.S. Pages 3/4 9-1 9-10 9-11 9-12

          .06/01/84                                                120                                                 Revision A L      - ._.            _  _    _ _ -   . _ _ _ _ _ _   . - _ _ _         .- . . -.               . _ _ __ __ _ __~ _ _

_ _m

j 1 TECHNICAL SPECIFICATIONS SELECTED RELEVANT REGULATIONS , Thie 10-Energy 9 50'.11 mined that there are no unresolved (2)(1) The processing. fabrication or anfety issues relating to the additional refining of special nuclear material or i activitie. tat may be authorized pur- the separation of.special nuclear mate. i suant tc .his -paragraph that would rial, or the separation of special nucle. ( l constitute good cause for withholding ar material from other substances by a l authorization. prime contractor of the Department l (4) Any activities undertaken pursu. under a prime contract for; i ant to an authorization granted under (A) The performance of work for the j this paragraph shall be entirely at the Department at a United States govern-risk of the applicant t nd, except as to 1 matters determined under paragraphs ment owned or controlled site: ? (e)(2) and (e)(3)(11), the grant of the (B) Research in, or development. authorization shall have no bearing on manufacture, storage testing or trans-the issuance of a construction permit portation of, atomic weapons or com- ( I with respect to the requirements of ponents thereof; or ; the Act, and rules, regulations, or (C) The use or operation of a pro- ! orders promulgated pursuant thereto, duction or utilization facility in a (Secs.101.185. 68 Stat. 936. 956, as amended United States owned vehicle or vessel;/ or l 10 3 tat 853 2 .SI432'); c." 2 1. (11) By a prime contractor or subcon. l amended. Pub .L 93-438. 88 Stat.1242. Pub. tractor of the Commission of the .De , L 94.~9. 89 Stat. 413 (42 U.S.C. 5841); sec. partment under a prime contract or ] 161 as at: ended. Pub. L 83-703. 68 Stat. 948 subcontract when the Commission de-(42 U.S.C. 22011) termines that the exemption of the (21 FR 355. Jan.19.1956, as amended at 25 prime contractor or subcontractor is FR 8712. Sept. 9.196o: 33 FR 2381, Jan. 31 authorized by law; and that, under the 1968; 35 FR 11460. July 7,1970; 37 FR 5748 Aiar.2.1.1972; 39 TR 14508. Ant. 24,1974: 39 - terms of the contract or subcontract.

                                       "PR 26279. July 18,1974; 39 TR 332o2. Sept.      there is adequate assurance that the 16.1974: 42 FR 22887. May 5.1977; 43 FR work thereunder can be accomplished 6924.Feb.17,1978)                                without undue risk to the public health and safety; 4 50.11 Exceptions and exemptions from              (2)(1) The construction or operation licensing requirements.

of a production or utilization facility Nothing in this part shall be deemed for the Department at a United States i to require a license for: government owned or controlled site, i (a) The manufacture, production or including the transportation of the acquisition by the Department of De- production or utilization fac,lity to or I fense of any utilization facility author- from such site and the performance of Ized pursuant to section 91 of the Act, contract services during temporary in-or the use of such facility by the De- terruptions of such transportation; or partment of Defense or by a person the construction or operation of a pro-under contract with and for the ac- duction or utilization facility for the count of the Department of Defense; Department in the performance of re-(b) Except to the extent that Admin- search in, or development. manufac' istration facilities of the types subject to licensing pursuant to section 202 of ture, storage testing, or transporta-the Energy Reorganization Act of tion of, atomic weapons or components - 1974 ' are involved; thereof; or the use or operation of a production or utilization facility for the Department in a United . States

                                           *The Department facilities identified in government owned vehicle or vessel:

1) De o$tation IJould Metal Past ** **

Breeder reactors when operated as part of ducted by a prime Contractor of the the poser reneration facilities of an electric - utility sys'em. of when operated in any 1975, when operated a.s part of the power other manner for the purpose of demon. generation faci!!tles of an electric utility strating the suitability for commercial ap. system or when operated in any other p!! cation of such a reactor. . manner fo.- the purpose of demonstrating (2) Other demonstrauon nuclear reactors. the suitability for commercial application of except those in existence on January 19. such a reactor. 392 . 06/01/84 121 Revision A

Chapter 1-Nuclear Regulatory Commission { 50.21 Department under a prime contract meet those needs on a timely basis and with the Department. delay costs to the applicant and to (11) The construction or operation of consumers. a production or utilization f acility by a - Issuance of such an exemption shall prime contractor or subcontractor of the Commission or the Department not be deemed to constitute a commit-ment to issue a construction permit, under his prime contract or subcon-tract when the Commission deter. During the period of any exemption ! mines that the exemption of the granted pursuant to this paragraph

                                    -prime contractor or subcontractor is (b), any activities conducted shall be authorized by law; and that, under the        carried out in such a manner as will terms of the contract or subcontract. minimize or reduce their environmen-there 1.s adequate assurance that the tal impact.

work thereunder can be accomplished (37 m 5748. Mar. 21.1972. a.s amended at without undue risk to the public 39 m 26279. July 18,1974; 40 m 8789. Mar. health and safety. 3.1975) (c) The transportation or possession of any production or utilization facili. I 50.13 Attacks and destructive act.s by en. ty by a common or contract carrier or emies of the United States; and defense warehousemen in the regular course activities. of carriage for another or storage inci* An applicant for a license to con. dent thereto, struct and operate a production or uti-(40 m 8788. Mar. 3.19751 lization facility, or for an amendment to such license, is not required to pro- - 8 50.12 Speelfic exemptions. vide for design features or other meas. (a) The Commission rnay, upon ap.' ures for the specific purpose of protec- i plication by any interested person or tion against the effects of (a) attacks ; upon its own initiative, grant such ex. and destructive acts, including sabo. emptions from the requirements of tage, directed against the facilhy by the regulations in this part as it deter. an enemy of the United States, wheth-

                                  ' mines' are authorized by law and will          er a foreign government or other not endanger !!fe or property or the          person, or (b) use or deployment of                                                   j common defense and security and~are           weapons incident to U.S. defense activ-                                               i otherwise in the public interest.           i ities.                                                                                l (b) Any person may request an ex-          (32 m 13445 Sept. 26,19671                                                            l emption perntitting the conduct of ac-tivities prior to the issuance of a con

CussIrtCAT!oN AND Dr.sCRIrrtoM or struction permit prohibited by i 50.10. LIcrNsts >

The Commission may grant such an - exemption upon considering and bal- t 50.20 Two classes of licenses. ancing the following factors: (1) Whether conduct of the proposed Licenses will be issued to named per-activities will give rise to a significant sons applying to the Commission adverse impact on the environment: therefor, and will be either class 104 or class 103, and the nature and extent of such. e e redress of any adverse 85M1 Class W Hunsa: for mdcal _ environment impact from conduct of therapy and research and development the proposed activities can reasonably I** U "* be effected should such redress be nec. A class 104 license will be issued, to essary; an applicant who qualifies, for any one (3) Whether conduct of the proposed or more of the following: to transfer or activities would foreclose subsequent receive in interstate commerce, manu. adoption of alternatives; and facture, produce, transfer, acquire. (4) The effect of delay in conducting possess, or use. such activities on the public interest, (a) A utilization facility for use in including the power needs to be used rnedical therapy; or by the proposed facility, the availabil- (b)(1) A production or utilization fa-ity of alternative sources, if any, to cility the construction or operation of 393 - 06/01/84 122 Revision A

 .                                                                                                                                                                   1 1

I l (4) The information described in minimum information

to be htcluded paragraphs (a)(1) and (2) of this sec- shall consist of the following: l tion shall be submitted as a separate (1) A description and safety assess-document prior to any other part of ment of the site on which the facility .

the license appliettion as provided in is to be located, with appropriate ,at- , I paragraph (b) and in accordance with tention to features affecting facility 4 2.101 of this chapter. design. Special attention should be di-(b) Except as provided in paragraph rected to the site evaluation factors (d), any person who applies for a class identified in Part 100 of this chapter. 103 construction permit for a nuclear . Such assessment shall contain an anal-power reactor on or af ter July 28,1975 ysis and evaluation of the major struc-shall submit the document titled "In* tures, systems and components of the formation Requested by the Attorney facility which bear significantly on the General for Antitrust Review" at least acceptability of the site under the site nine (9) months but not more than evaluation factors identified in Part thirty six months prior to the date of 100 of this chapter, assuming that the submittal of any part of the applica-facility will be operated at the ulti-tion for a class 103 construction permit, mate power level which is contemplat- l ed by the applicant With respect to ' (c) (Reservedl operation at IJie projected initial ' (d) Any person who applies for a class 103 construction permit for a nu- power level, the applicant is required clear power reactor pursuant to the to submit information prescribed in provisions of 12.101(a-1) and Subpart paragraphs (a)(2) through (8) of this l F of Part 2 of this chapter shall g g gg g ) submit the document title "Informa- quired by this paragraph. In support i j tion Requested by the Attorney Gen of the application for a construction j eral for Antitrust Review" at least permit. ;

     '                      nine (9) months but not more than                (2) A summary description and dis-                    ,                                 ;

thirty six months prior to the filing of cussion of the facility, with special at- l part two or part three of the applica. tention to design and operating char- l tion, whichever part it. filed first, as acteristics, unusual or novel design ! specified in { 2.101(a-1) of this chap- features, and principal safety consider-ter, ations. 1 (e) Any person who applies for a (3) The preliminary design of the f a- i l class 103 construction permit for a clltty including: uranium enrichment or fuel reprocess-l (1) The principal design criteria for i ing plant shall submit such,informa- the f acility.* Appendix A. General . l tion as may be requested by the Attor* Design Criteria for Nuclear Power i ney General for antitrust review, as a Plants, establishes minimum require-separate document as soon as possible ments for the principal design crittria and in accordance with 12.101 of this for water cooled nuclear power plants chapter, similar in design and location to plants (Sec.102. Pub. L 91-190. 83 Stat. 853 (42 for which construction permits have U.S.C. 4332); sec. 201, as amended. Pub. L previously been issued by the Commis- _4 93-438. 86 Stat.1242. Pub. L 94-79,89 Stat. sjon and provides guidance to appli-l 413 (42 U.S.C. 58411) cants for construction permits in es- I l (39 FR 34395. Sept. 25.1974 as amended at tablishing principal design criteria for 42 FR 22887. 19.1977; 43 FRMay S.1977; 49775, 42 FR 25721, Oct. 25.1978; 44 FR May other types of nuclear power unitsi 60716. Oct. 22,19791. >.

                                                                             'The applicant may provide Information Il $0.34 Contents of applications; technleal required by this parsgraph in the form of a Information.                            discussion, with specific references of simi.

j larltles to and differences from, facilities of (a) Preliminary safety analysis similar design for which applications have report. Each application for a con. previously been filed with the Commission. struction permit shall include a pre, eGeneral design criteria for chemical liminary safety. analysis report. The processing f acilities are being developed. 399

          .06/01/84                                                     123                                                Revision A

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                        $ 50.34 Title 10-Energy              '

(11) The design bases and the rela- the quality assurance program for a lion of the design bases to the princi- nuclear power plant or a fuel repro- j pal design critcria; cessing plant shallinclude a discussion . (iii) Inicrmation relative f.o materi- of how the ap;licable requirements of ] als of construction. general arrange. Appendix B will be satirfled. ; d ment, and approximate dimensions. (8) An identification of those struc-sufficient 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-margin for safety. quacy of their design; and identifica-(4) A preliminary analvsis and evala- tion and description of the research ation of the design and performance and development program which will of structures, systems, and compo- be conducted to resolve any safety nents of the facility with the objective questions associated with such struc-of assessing the risk to public health tures, systems or components; and a and safety resulting from operation of schedule of the research and develop-the facility and including determina- men t. program showing that such Lion of (1) the margins of safety during safety questions will be resolved at or normal operations and transient con % before the latest date stated in the ap-tions anticipated during the life of the plication for completion of construc-f acility, and (11) the adequacy of struc- tion of the facility. tures, systems, and components pro- (9) The technical qualifications of vided for the prevention of accidents the applicant to engage in the pro-and the mitigation of the conse- posed activities in accordance with the quences of accidents. Analysis and regulations in this chapter. evaluation of ECCS coollng perform- (10) A discussion of the applicant's ance following postulated loss of 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 shall be included in these i 50.46 of this part for facilities for plans. . which construction permits may be (11) On or af ter February 5,19'l9, issued af ter December 28. 19'14. applicants who apply for cont.truction (5) An identification and justifica- permits for nuclear powerplants to be tion for the selection of those varia- built on multlunit sites shall identify bles, conditions, or other items which potential hazards to the structures, are determined as the result of pre- systems and components important to Ilminary safety analysis and evalua- safety of operating nuclear facilities tion to be probable subjects of techni- from construction activities. A discus-cal specifications for the facility, with slon shall also be included of any man-special attention given to those items agerial and administrative controls which may significantly influence the that will be used during construction final design: Provided, hotcerer That to assure the safety of the operating this requirement is not applicable to unit. an application for a construction (b) Final sc/ety analysis repo rt. permit filed prior to January 16,1969. Each application 'or a license to oper-(6) A preliminary plan for the appli- ate a facility shall include a final cant's organization, training of person- safety analysis report. The final safety - nel, and conduct of operations. analysis report shall include informa-(7) A description of the quality as- tion that describes the facility, pre-surance program to be applied to the sents the design bases and the limits design, fabrication, construction, and on its operation, and presents a safety testing of the structures, systems, and analysis of the structures, systems, components of the facility. Appendix and components and of the facility as B " Quality Assurance Criteria for Nu- a whole, and shall include the follow-clear Power Plants and Fuel Repro- Ing: cessing Plants," sets forth the require- (1) All current Information, such as ments for quality assurance programs the results of environmental and me-for nuclear power plants and fuel re- teorological monitoring programs, processing plants. The description of which has been developed since lasu- 2 400 i ) , I I I 06/01/84 124 Revision A i

m 6 l l Chapter b-Nuclear Itegulatory Commission 6 50.34 ance of the construction permit, relat- (5) A desertption and evaluation of Ing to site evaluation factors identified the results of the applicant's pro-in Part 100 of this chapter, grams, including research and develop-(2) A description and analysis of the ment, if any, to demonstrate that any structures, systems, and components safety questions identified at the con. of the facility, with emphasis upon struction permit stage have been re-performance requirements. the bases, solved. { with technical justification therefor, (6) The following information con-upon which such requirements have cerning facility operation: been established, and the evaluations (1) The applicant's organizational required to show that safety functions structure, allocations or responsibil-will be accomplished. The description ities and authorities, and personnel shall be sufficient to permit under, qualifications requirements. standing of the system designs and (11) Managerial and administrative their relationship to safety evalua. ' controls to be used to assure safe oper-tions. ation. Appendix B, Quality Assurance (i) For nuclear reactors, such items Criteria for Nuclear Power Plants and es the reactor core, reactor coolant Fuel Reprocessing Plants, sets forth l system, instrumentation and control the requirements for such controls for j systems, electrical systems, contain- nuclear power plants and fuel repro-ment system, other engineered safety cessing plants. The information on the ) features, auxiliary and emergency sys- controls to be used f or a nuclear power ' tems, power conversion systems, radio- plant or a fuel reprocessing plant shall active waste handling systems, and include a discussion of how the appll-fuel handling systems shall be dis- cable requirements of Appendix B will cussed insofar as they are pertinent. I W) For. facilities other than nuclear N(118f) Ia for preoperationa! testing reactors, such items as the chemical, and bitial operations. physical, metallurgical, or nuclear (iv) Plans for conduct of normal op-process to be performed, instruments- erations, including maintenance, sur-tion and control systems, ventilation veillance, and periodic testing of struc- i and filter systems. electrical systems tures, systems, and components. ; auxiliary and emergency systems, and' (v) Plans for coping with emergen-cies, which shall include the items ] radioactive waste handling systems i shall be discussed insofar as they are specified in Appendix E. ! perunent. (vi) Proposed technical specifications l prepared in accordance with the re I (3) The kinds and quantities of ra- quirements of i 50.36. dioactive materials expected to be pro- (vil) On or after February 5,1979, duced in the operation and the-means for controlling and limiting radioactive app!! censes cantsforwho applypowerplants nuclear for operating to11-be effluents and radiation exposures operated on multfunit sites shall in-within the !!mits set forth in Part 20 clude an evaluation of the potential of this chapter- hazards to the structures, systems, and - (4) A fina.l a.n,talysis and evaluation of compon' ents important to safety of op- ; l the design and performance of strue. erating units resulting from construc- 1 tures, systems, and components with tion activities, as well as a description the objective stated in paragraph 'I of the managerial and administrative ! (a)(4) of this section jnGakinLinto controls to be used to provide assur- l account any certinent :nfortnation de- ance that the !!miting conditions for l veloped since the subm ttaT~of the pre- operation are not exceeded as a result liminary safety analysis report. Analy- of construction activities at the mul-sis and evaluation of ECCS cooling tiunit sites, performance following postulated loss- (7) The technical quallflettlons of of-coolant accidents shall be per- the applicant to engage in the pro- , formed in accordance with the re- posed activities in accordance with the quirements of 150.46 for facilities for regulations in this chapter, which a license to operate may be (8) A description and plans for im-issued after December 28,1974. plementation of an operator requalifi-401 06/01/84 , 125 Revision A k

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Title 10-Energy . j { 50.34a connected to the containment atmos- tions thereof, that underlie the corre. phere. (ll.E.4.1) sponding SRP acceptance criteria. (vil) Prcvide a description of the (3) The SRP was issued to establish management plan for design and con- criteria that the NRC staff intends to struction activitics, to include: ( A) The use in evaluating whether an appil-organizational and management strue- cant /1lcensee meets the Commission's ture singularly responsible for direc. regulations. The SRP la not a substi. l tion of design and construction of the tute for the regulations, and compil-proposed plant; (B) technical re- ance la not a requirement Applicants sources director by the applicant; (C) shall identify differences from the l details of the interaction of design and SRP acceptance criteria and evaluate l construction within the applicant's or- how the proposed alternatives to the l ganization and the manner by which SRP criteria provide an acceptable i the applicant will ensure close integra- ~ method of complying with the Com- I tion of the architect engineer and the mission's regulations. nuclear steam supply vendor; (D) pro- (Secs.161b 1611. Pub. L $3 703, 68 Stat. posed procedures for handling the s 2 2 1) 4 8 transition to operation; (E) f.he degree g {2 2 1, of top level management oversight and 5844); sec. 7. Pub. L 93-377. 88 Stat. 475: technical control to be exercised by sec.1811. Pub. L 83 703, 68 Stat 948 (42 1 the applicant during design and con. U.S.C. 2201)) struction. including the preparation (33 FR 18612. Dec.17.1968, as amended at l and implementation ' of procedures 34 FR 6037. Apr 3.1969; 34 FR 6770. Apr. I l necessary to guide the *!ffort. (II.J.3.1) 23.1969; 35 PR 1o499. June 27.1970; 35 FR i (g) Conformance tof tA the Standard 19567, Dec. 24,1970; 36 FR 3256. Feb. 20. . Revieto Plan (SRP). (1)(l) Applications - 1971; 36 FR 4861. Mar.13,1911; 36 FR ! l for light water cooled nuclear power 18201. Sept.11.19711 plant operating licenses docketed after EortontA1. Nort- For additional PtnenAt. May 17,1982 shall include an evalua- Rectstra citations affecting I 50.34 see the , I tion of the facility against the Stand. List of CPR sections Affected in the Finding ard Review Plan (SRP) in effect on Aids section of this volume. l May 17,1982 or the SRP revision in I $0.34a Design objectives for equipment effect six months prior to the docket t control releases of radioactive mate-date of the application, whichever is risi in emuenwnuclear pown nac. later, (11) Applications for light water I' ? . cooled nuclear power plant construc- (a) An app!! cation for a permit to tion permits, manufacturing licenses, construct a nuclear power - reactor and preliminary or final design appro- shall include a description of the pre-vals for standard plants docketed af ter liminary design of equipment to be in-May 17,1982 shall include an evalua. stalled to maintain control over radio . tion of the facility against the SRP in active materials in gaseous and liquid effect on May 17, 1982 or the SRP re- - effluenta produced during normal re-vision in effect six months prior to the actor operations, includlng expected docket date of the application, which- operational occurrences, In the case of ever is later. . . an application filed on or after Janu-(2) The evaluation required by this ary 2.1971, the application shall also - section shall include an identification Identify the design objectives, and the and description of all differences in means to be employed. for- keeping design 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. la reasonably achievable.The term "as tures, techniques, and measures given low as is reasonably achievable" as in the SRP acceptance criteria. Where used in this part means as low as is such a difference exists. the evalua- reasonably achievable taking into ac-tion shall discuss how the alternative count the state of technology, and the proposed provides an acceptable economics of improvement in relation method of complying with those rules to benefits to the public health and. or regulations of Commission, or por- safety and other societal and socioeco-408 06/01/84 126 Revision A

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   <  s                                                                                                                                                                                                          ,1
           ,                                                                                                                                                                                                              j
        ,                                  pgLu                                                                                                                     Tifle 10-Energy                                     J (b) A construction permit will consti- tions. The technteal specifications will tute an authorization to the appliennt be derived from the analvscs and aval-                                                                                               '1 to proceed with construction but will untion included in the safety analysis l                                         not constitute Commission approval of                                                        re  on, ano amenqn, tenM thereto. sub-the safety of any design feature or m e pursuant to,,) 50.34. The Com-                                                                                              "           y j

specification unless the applicant spe- mission may lMlude such additional j cifically requests such approval and . technical specifications as the Com-such approval is incorporated in the mission finds appropriate. . ] permit. The applicant. at his option. (c) Technical specifications will in. may request such approvals in the clude items in the following categories: j 3 construct!on permit or, from. time to (1) Safety limit.t. limiting 44/ety j time, by amendment of his construc- system settings, and limiting control J tion permit. The Commission may, in . settings. (1)(A) Safety limits for nucle- l Its discretion, incorporate in any con- ' ar reactors are limits upon important ' . struction permit provisions requiring process variables which are found to i the appilcant to furnish periodic re- be necessary to reasonably protect the .) ports of the progress and results of re- j integrity of certain of the physical -1 search and development programs de* barriers which guard against the un. I signed to resolve safety questions. controlled release of radioactivity. If (c) Any construction permit will be any safety limit is exceeded, the reac-subject to the limitation that a license tor shall be shut down. The licensee authorizing operation of the facility . shall notify the Commission, review will not be issued by the Commission the matter and record.the results of until (1) the applicant has submitted the review, including the cause of the to the Commission, by amendraent to condition and the basis for corrective 1 the application, the complete final action taken to preclude reoccurrence. safety analysis report, portions of Operation shall not be resumed until I I

which.may be submitted and evaluat. authorized by the Commission. '

ed from time to time, and (2) the Com. (B) Safety limits for fuel reprocess. I mission has found that the final ing plants are those bounds within .I design provides reasonable assurance which the process variabics must be i that the health and safety of the maintained for adequate control of public will not be endangered by oper- the operation and which must not be ation of the facility in accordance with exceeded in order to protect the integ- i the requirements of the license and rity of the physical system which is ' the regulations in this chapter. designed to guard against the uncon-(Sec.185,68 Stat. 955; 42 U.S.C. 2'235) trolled release of radioactivity. If any [27 m 12915. Dec. 29,1962. a.s arnended at safety limit for a fuel reprocessing 31 m 12780, sept. 30.1986; 35 m 5318, plant is exceeded, corrective action Mar. 31,1970; 35 m 6644. Apr. 25,1970; 35 shall be taken as stated in the technl. m 11461. July 7.19701 cal specification or the affected part 1 50.36 Technical speelrications, of the process, or the entire process if required, shall be shut down, unless (a) Each applicant for a license such action would further reduce the authorizing operation of a production margin of safety. The licensee shall or utilization facility shall include in notify the Commission, review the ' his application proposed technical matter and record the results of the - specifications in accordance with the review, including the cause of the con-requirements of this section. A sum. dition and the basis for corrective mary statement of the bases or rea; action taken to preclude reoccurrence. sons for such specifications, other, If a portion of the process or the i than those covering administrative entire process has been shut down, op-controls, shall also be included in the erstion shall not be resumed until au-application, but shall not become part thorized by the Commission. of the technical . specifications. (ll)(A) Limiting safety system set-(b) Each license authorizing oper. tings for nuclear reactors are settings atton of a production or utilization fa. for automatic protective devices relat. cility of a type described in I 50.21 or ed to those variables having signifi-150.22 will include technleal speciflCA- Cant safety functions. Where a limit-410 06/01/84 127 Revision A C__._____i______ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ . _ . _ _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ . _ _ _ . . _ . _ . _ _ _ . . _ . . _ _ i___.

Chapter 1-Nuclear Regulatory Commission @ 50,36 . Ing safety system setting is specified any remedial action permitted by the for a variable on which a safety limit technical specification until the condi. has been placed, the setting shall be so/ tion can be met. In the cue of either a chosen that ytnmt_Ic protect!"e nuclear reactor or a fuel reprocessing action will correct the abnormal situa- plant. the licensee shall notify the tion before a cafety limit is exceeded. Commission, review the matter, and If, during operation, the automatic record the results of the review, in-safety system does not function as re- cluding the cause of the condition and quired, the licensee shall take appro- the basis for corrective action taken to priate action, which may include M1pt- preclude reoccurrence. tingdown the reactor. He shall notify (3) SurveUlance requirements Sur. the Commission, review the matter veillance requirements are require-and record the results of the review. ments relating to test, calibration, or including the cause of the condition inspection to assure that the necessary and the basis for corrective action quality cf systems and components is

taken to preclude reoccurrence, maintained that facility operation will (B) Limiting control settings for fuel be within the safety limit.s. and that reprocessing plants are settings for the limiting conditions of operation automatic alarm or protective devices will be met, related to those variables haing sig- (4) Design katures. Design features nificant safety functions. Where a to be included are those features of II,miting control setting is specified for the facility such as materials of con-a variable on which a safety lirait has struction and geometric arrangements, been placed, the setting shall be so which, if altered or modified, would chosen that protective action, either have a significant effect on safety and automatic or manual..will correct the are not covered in categories described abnormal situation before a safety .in paragraphs (c) (1), (2), and (3) of limit is exceeded. If, durin-g operation, this secti the automatic alarm or protective d2-vices do not function as required, the istrative controls are the provisions re-licensee shall take appropriate action to maintain the variables within the lating to organization and manage, limiting control setting values and to ment, procedures, recordkeeping, review and audit, and reporting neces. repair promptly the automatic devices or to shut down the affected part of sary to assure operation of the facility in a safe manner, the process and, if required, to shut down the entire process for repair of (d)(1) This section shall not be automatic devices. The licensee shall deemed to modify the technical spect-notify the Comminich, review the fications included in any license issued matter, and record the results of the p. lor to January 16,1969. A license in review, including the cause of the con. which technical specifications have dition and the basis for corrective not been designated shall be deemed action taken to preclude reoccurrence, to include the entire safety analysis (2) Limiting condiffons for oper. report as technical specifications, affon. Limiting conditions for oper. , (2) An applicant for a license author-ation are the lowest functional c,gne . Izing operation of a production or uti. bility,.2r performance Tevels of equJ . lization facility to whom a construe. _ ment required for safe opeM6n o tion permit has been issued prior to thii TiIcility. When a limiting conditio January 16, 1969, may submit techni. for operation of a nuclear reactor is cal specifications in accordance with not met, the !!censee shall shut down this section, or in accordance with the the reactor or follow any remedla! requirements of this part in effect action permitted by the technical spec! prior to January 16,1969. (fication until the condition can be ' (3) At the initiative of the Commis. met When a limiting condition for op- slon or the licensee, any license may eration of any process step in the be amended to include technical speci- , system of a fuel reprocessing plant is i fications of the scope and content l not met, the licensee shall shut down which would be required if a new 11- l that part of the operation or follow cense were being issued, I 411 1 06/01/84 128 Revision A J

-v - Chapter I-Nuclear Regulatory Commi .lon 9 50 #2 5 50.3R Ineligibility of certain applicanta. (d) Any applienble requirements of ; Any person who is a citizen, nation- Part 51 have been satisfied. l al, or agent of a foreign country, or (21 FR 355. Jan.19.1956. as amended at 16 l any corporation, or other entity which FR 12731. July 7.1971; 39 FR 26219. July j the Commission knows or has reason 18.1974: 47 FR 13754. Mar. 31,19821 l to believe is owned, controlled, or dominated by an allen, a foreign cor. 8 50.41 Additional standards for clana 104 poration, or a foreign government. liC'"*- shall. be ineligible to apply for and In determining that a class 10411 obtain a license, cense will be issued to an appliennt, (Sec.161 as amended. Pub. L 83-703. 68 the Commission will, in additio7 to ap-Stat. 948 (42 U.S.C. 22o1); sec. 201, as plying the standards set forth in 3 amended. Pu.b. L 93 438, 88 Stat.1243 (42 150.40 be guided by the followira ;on. I U.S.C. 5841)) siderations: l 121 FR 355. Jan.16. J56, as amended at 43 (a) The Commission will permit the l FR 6924. Feb.17,197 tl widest amount of effective medical ! therapy. possible with the amount of I 8 50.39 Public inspection of application, l special nuclear material available for Applications and documents submit- such purposes.

ted to the Commission in connection (b) The Commission will permit the ' with applications may be made avalla- conduct of widespread and diverse re- i ble for public inspection in accordance search and development. with the provisions of the regulations (c) An application for n class 104 op-contained in Part 2 of this chapter, crating license as to which a person who intervened or sought by timely STANDARDS roR LICzNSEs AND CoNsrauc11oM Pznuzzs written notice to the Commission in intervene in the construction permit ! 8 50.40 Common standarda. proceeding for the facility to obtain a determination of antitrust consider-In determining that a license will be ations or to advance a jurisdictional issued to an applicant, the Commis- basis for such determination has re-slon will be guided by the following quested an antitrust review under sec-considerations: tion 105 of the Act within 25 days (a) The processes to be performed,' after the date.,of publication in the the operating procedures, the facility l FEncaA1. Rzo!sTra of notice of filing of l and equipment, the use of the facility, the application for an operatinE 11- ' and other technical specifications, or cense or December 19,1970, whichever the proposals, in regard to any of the is later, is also subject to the provi-foregoing collectively provide reason-able assurance that the applicant will sions of i 50.42(b). I i comply with the regulations in this (42 U.S.C. 2132-2135,2239) chapter, including the regulations in 121 FR 355 Jan.19,1956. as amended at 35 Part 20, and that the health and FR 19660, Dec. 29,19701 l safet of the public will not be endan-8 504

Additional standards for class 103 (b) The applicant is technically and U"*'*- _

financially quallfled to engage in the In determining whether a class 103 proposed activitles in accordance with license will be issued to an applicant, the regulations in this chapter. How. the Commission will, in addition to ap-ever, no consideration of financial plying the standards set forth in qualifications is necessary for an elec. I 50.40, be guided by the following tric utility applicant for a license for a considerations: i production or utilization facility of the (a) The proposed activities will serve { type descrit,ed in 150.21(b) or 150.22. a useful purpose proportionate to the ' (c) The issuance of a license to the quantitles of special nuclear material applicant will not, in the opinion of or source material to be utilized. the Commission, be inimical to the (b) Due account will be taken of the common defense and security or to the advice provided by the Attorney Gen- i health and safety of t.he public, eral, pursuant to subsection 105c of i l 413 l 1

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06/01/o4 129 Revision A

w- , , L 4 Chapter l-Nucteer Regulatory Commission { 50.46 the general requirements of Criteria (2) A combustible gas control system

41. 42, and 43 of Appendix A to this is a system that operates after a LOCA part. If a purge system is used as part to maintain the concentrations of, of the repressurization system, the combustible gases within the contain-purge system shall be designed to con- ment, such as hydrogen below flam-form with the general requirements of mability limits. Combustible gas con-Criteria 41. 42, and 43 of Appendix A trol systems are of two types: (1) Sys-to this part. The containment shall tems . that allow controlled release not oe repressurized beyond 50 per. from containment, through filters if cent of the containment design preg. necessary, such as purging systems ,

and repressurization systems, and (ii) sure. (g) For facilities with respect to systems that do not result in a signifi-which the notice of hearing on the ap- cant release from containment such as , plication for a construction permit was mombiners. q published on or before December 22. -(3) A purging system is a system for ' 1968. if the combined radiation dose at the controlled release of the contain-the low population zone outer bound, ment atmosphere to the environment through filters if needed. try from purging (and repressuriza. (4) A repressurization system is a tion if a repressurization system is pro- system used to dilute the concentra-vided) and the postulated LOCA calcu- tion of combustible gas within contain- . lated in accordance with 1 100,11(a)(2) ment by adding' inert gas or air to the l of this chapter is less than 25 rem to containment. Dilution of the combus- 1 the whole body and less than 300 rem tible gas results in a delay in time .j I to the thyroid, only a purging system la necessary, provided that the purging tm a Damma e . comentradon is - 1 system and any filtration system asso- reache,d and permits fission -! decay. Operation is limited to product a con-clated with it are designed to conform tainment repressurization to 50 per-with the general requirements of Cri- cent of the containment design pres-teria 41. 42, and 43 of Appendix A to sure. A purging system is normally this part. Otherwise, the facility shall part i the repressurization system. be provided with another type of com-bustible gas control system (a repres. (sec.161, as amended. Pub. L. 83-703. 68 surization system is acceptable) de. Stat. 948 H2 U.S.C. 2201% sec. 201, as i signed to conform with the general re. Lamended. Pub. L. 93-433. 88 Stat.1242. Pub. - q 9679, se stat. 413 H2 Est 5842n quirements of Criteria 41,42, and 43 of ] Appendix A to this part. If a purge (43 m 50163. Oct. 27,1978. as amended at , system is used as part of the repressur- 46 m SM. N 2. 280 1tation system, it shall be designed to conform with the general require. 5 50.45 Standards for construction per. mits. l ments of Criteria 41. 42, and 43 of Ap. l pendix A to this part. The contain- An applicant for a license or an ment shall not be repressurfzed amendment of a license who proposes j beyond 50 percent of the containment to construct or alter a production or design pressure, utilization facility ' will be initially (h) As used in this section: (1) Deg- granted a construction permit if the radation, but not total failure. ' of application is in conformity with and - emergency core cooling functioning acceptable under - the criteria of j means that the performance of the . Il 50.31 through 50.38 and the stand-emergency core cooling system is pos- ards of Il 50.40 through 50.43, tulated, for purposes of design of the combustible gas control system not to 8 50A6 Acceptance criteria for emergency meet the acceptance criteria in i 50.46 core cooling systems for light water , and that there could be localized clad nuclear Pom wtors. melting and metal water reaction to (axi) Except as provided in para-the extent postulated in paragraph (d) graph (aK2) and (3) of this section, of this section. The degree of perform- each boiling and pressurized light-ance degradation is not postulated to water nuclear power reactor fueled be sufficient to cause core meltdown. with uranium oxide pellets within cy. 417 , 06/01/84 130 Revision A L - - - - - - _ _ _ _ - - - _

4 f 50A6 Title 10 g lindrical 7,trealoy cladding shall be complete it. The Director of Regula-provided with an emergency core cool- tion of the Atomic Energy Commisuon ing system (ECCS) which shall be de-signed such that its calculated cooling shall have caused notice of such a re-quest to be published promptly m the performance following postulated loss. FZDERAL RecIstra: such notice snall l of coolant accidents conforms to the have provided for the submission of l criteria set forth in paragraph (b) of comments by interested persons l this section. ECCS cooling perform- within a time period established by I ance shall be calculated in accordance the Director of Regulation. If, upon I with an acceptable evaluation model, reviewing the fomgoing submissions. l and shall be calculated for a number the Director of Regulation conduded l of postulated loss of-coolant accidents that good cause had been shown for l of different sizes, kcations, and other an extension, be may have extended i properties sufficient to provide assur- the six month period for the shortest i ance that the entire spectrum of pos- additional time which in his judgmem tulated loss of coolant accidents is cov- will be necessary to enable the licensee f i ered. Appendix K, ECCS Evaluation to furnish the submissions required by ) Models, sets forth certain required and paragraph (aX2X11) of this section. Re- i acceptable features of evaluation questa for extensions of the six month l models. Conformance with the criteria period submitted under this subpara. i set forth in paragraph (b) of this sec. graph will have been ruled upon by l tion with ECCS cooling performance the Director of Regulation prior to ex-calculated in accordance with an ac. piration of that period. ceptable evaluation model, may re. (iv) Upon submission of the evalua. quire that restrictions be imposed on tion required by paragraph (aM2X1il of reactor operation, this section (or under paragraph (2)'Wtth respect to reactors for (aX 2 Xfil), if the six month period is which operating licenses have previ- extended) the facility shall continue ously been issued and for which oper- or commence operation only within ating licenses may issue on or before the limits of both the proposed techni-December 28,1974: cal specifications or license amend-(1) The time within which actions re- Inents submitted in accordance with quired or permitted under this para- this paragraph (aX2) and all technical graph (ax2) must occur shall begin to specifications or license conditions run on February 4,1974, previously imposed by the Atomic (11) Within six months following the Energy Commission, including the re-date specified in paragraph (aX2XD of quirements of the Interim Policy this section an evaluation in accord. Statement (June 29, 1971, 36 FR ance with paragraph (aX1) of this sec- 12248)-as amended December 18,1971, tion shall have been submitted to the 36 FR 24082). Director of Regulation of the Atomic (v) Further restrictions on reactor Energy Commission. The evaluation operation will be imposed if it is found shall have been accompanied by such .: at the evalua(lons submitted under proposed changes in technical speciff. pacagraphs (ax2) (ID and (iii) of this cations or license amendments as may section are not cons! stent with para-be necessary to bring reactor oper. graph (aX1) of this section and as a - ation in conformity with paragraph result such restrictions are required to (aM1) of this section. protect the public health and safety, (111) Any licensee may have request- (vi) Exemptions from the operating ed an extension of the six month requirements of paragraph (aX2Xiv) period referred to in paragraph of this section may be granted for (aH2Xil) of this section for good cause. good cause. Requests for such exernp. Any such request shall have been sub- tion shall be submitted not less than mltted not less than 45 days prior to 45 days prior to the da.te upon which expiration of the six month period, the plant would otherwise be required and shall have been accompanied by to operate in accordance with the pro-affidavits showing precisely why the cedures of said paragraph (aX2Xiv) of evaluation is not complete and the this section. Any such request shall be minimum time believed necessary to filed with the Secretary of the Com-418 i 3 l i 06/01/84 131 Revision A ; f

v Chapter 1-Nuclear Regulatory Cornmission 9 50.46 mission, who shall cause notice of its ed to occur, the inside surfaces of the receipt to be published promptly in cladding shall be included in the oxi. the FEDERAL rec! STER, such notice dation, beginning at the calculated shall provide for the submission of time of rupture. Cladding thickness comments by interested persons before oxidation means the radial dis- ( within 14 days following FEDERA1. rec- tance from inside to outside the clad-IsTER publication. The Director of Nu- ding, af ter any calculated rupture or clear Reactor Regulation shall submit swelling has occurred but before sig-his views as to any requested exemp- nificant oxidation. Where the calculat. tion within five days following expira- ed conditions of transient pressure and l tion of the comment period. temperature lead to a prediction of l j (vil) Any request for an exemption cladding swelling, with or without submitted under paragraph (a)(2)(vi) cladding rupture, the unoxidized clad- l of this section must show, with appro- ding thickness shall be defined as the j l priate affidavits and technical submis- cladding cross sectional area, taken at sions, that it would be in the public in- a horizontal plane at the elevation of 1 terest to allow the licensee a specified the rupture, if it occurs, or at the ele-additional period of time within which vation of the highest cladding tem-to alter the operation of the facility in perature if no rupture is calculated to the manner required by paragraph occur, divided by the Sverage circum- 1 (aP? viv) of this section. The request ference at that elevation. For ruptured shall also include a discussion of the cladding the circumference does not l alternatives available for establishing include the rupture opening. l compliance with the rule, (3) Maximum hydrogen generation. (3) Construction permits may have The calculated total amount of hydro-been issued after December 28, 1973 gen generated from the chemical reac-but t efore December 28, 1974 subject tion of the cladding with water or l to any applicable conditions or restric- steam shall not exceed 0.01 times the tions imposed pursuant to other regu- hypothetical amount that would be lations in this chapter and the Interim generated if all of the metal in the Acceptance Criteria for Emergency cladding cylinders surrounding the l Core Cooling Systems published on fuel, excluding the cladding surro'inc'- June 29,1971 (36 FR 12248) as amend- ing the plenum volume, were to react. ed (December 18,1971, 36 FR 24082): (4) Coolable geometry. Calculated . l Provided, hotoever, that no operating changes in core geometry shall be license shall be issued for facilities such that the core remains amenable constructed in accordance with con- to cooling, struction permits issued pursuant to ($) Long term cooling. After any cal-this paragraph, unless the Commission culated successful initial operation of determines, among other things ~that the ECCS, the calculated core tem-the proposed facility meets the re- perature shall be maintained at an ac-quirements of paragraph (a)(1) of this ceptably low value and decay heat section. shall be removed for the extended (b )( 1) Peak cladding temperature. period of time required by the long-The calculated maximum fuel element lived radioactivity remaining in the l cladding temperature shall not exceed core. i 2200' F. (c) As used in this section: (1) Loss. - (2) Afarimum cladding oxidation. of coolant accidents (LOCA's) are hy. The calculated total oxidation of the pothetical accidents that would result l cladd!ng shall nowhere exceed 0.17 from the loss of reactor coolant, at a times the total cladding thickness rate in excess of trae capability of the i ! before oxidation. As used in this sub- reactor coolant makeup system, from

paragraph total oxidation means the brealts in pipes in the reactor coolant I

total thickness of cladding metal that pressure boundary up to and including I would be locally converted to oxide if a break equivalent in size 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 reactor coolant system. verted to stoichiometric zircon!um (2) An evaluation model is the calcu-dioxide. I.f cladding rupture is calculat- lational framework for evaluating the 419 06/01/84 132 Revision A

    ~                                                                             ,

l l

                                                                          .                                                 i ggw                                                                   Tiele 10 E-si behavior of the reactor system during       finding will constitute a rebuttable
                                                                                                                         'j a postulated loss of coolant accident        presumption on questions of adequacy                        .

(LOCA). It includes one or more com- and implementation capability. Emer-puter prorrams and all other informa- gency preparedness exercises (required 1 l tion ny mary for application of the by paragraph (b)(14) of this section calculational framework to a specific and Appendix E. Section F of this - LOCA. such as mathematical models part) are part of the operational in-used, assumptions included in the pro- spection process and are not required grams, procedure for treating the pro- for any initial licensing decision. gram input and output !nformation, (b) The onsite and, except as pro-specification of those portions of anal

vided in paragraph (d) of this section, ysia not included in computer pro- offsite emergency response plans for grams, values of parameters, and all nuclear power reactors must meet the other information necessary to specifF following standards:'

the calculational procedure. (1) Primary responsibilities for emer< (d) The requirements of this section sency response by the nuclear facility are in addition to any other require- licensee and by State and local organi- , ments applicable to ECCS set forth in zations within the Emergency Plan- ! this part. The criteria set forth in ning Zones have been assigned. *he . ! paragraph (b). with cooling perform- ; emergency responsibilities of the var-ance calculated in accordance with an tous supporting organizations have acceptable evaluation model, are in ! implementation of the general re-been specifically established, and each principal response organization has quirements with respect to ECCS cool- staff to respond and to augment its ing performance design set forth in pa initial response on a continuous basis. u at N 9[ kh -(2).On shift facility licensee respon-sibilities for emergency tesponse are (39 FR 1002, Jan. 4,1974. as amended at 39 unambiguously defined. Adequate l FR 27121. July 25,1974; 40 FR 8789. Mar. 3, staffing to provide initial facility acci- ! 19751 dent response in key functional areas S 50.47 Emergency plans. is maintained at all times, timely sug-mentation o:f response capabilities is (a)(1) Except as provided -in para- available and the interfaces among graph (d) of this section, no operating various onsite response. activities and license for a nuclear power reactor will- offsite support and response activities . be issued unless a finding is made by are specified. 1 NRC that there is reasonable ssaur- (3) Arrangements for requesting and ance that adequate protective meas-effectively using assistance resources j ures can and will be taken in the event have been made, arrangements to ac-

of a radiological emergency. i (2) The NRC will base its finding on commodate State and local staff at the a review of the Federal Emergency licensee's near site -Emergency Oper-ations Facility have been made, and Manageraent Agency (FEMA) findings other organizations capable of sug-and determinations as to whether menting the planned response have State and local emergency plans are been identified, adequate and whether there is reason.

able assurance that they can be imple- (4) A standard emergency classifica- - mented, and on the NRC assessment- tion and action level scheme, the bases as to .whether the applicant's onsite of which include facility system and emergency plans are - adequate and effluent parameters, is in use by the whether there is reasonable assurance nuclear fac!1 ty licensee, and Str.te and - that they can be implemented. A local re.sponse plans call for reliance i I FEMA finding will primarily be based on a review of the plans. Any other in. 'These standards are addressed by specif. formation already available to FEMA k crituta in NUREO-o650 FEM-REP 1 may be considered in assessing wheth- enutled criteria for Preparation and Eval-usuon of Radiological Emergency Response ! er there is reasonable assurance that Plans and Preparedness in support of Nucle. the plans can be implemented. In any ar Power Planta-for Interim Use and Com. NRC licer.ain4r proceeding, a FEMA ment", January 1980. 420 06/01/84 133 Revision A

                                                                                                                            ,      i, i

1 Chaptor I-Nuclear Regulatory Commission 9 5C.57 authorized by the Commission upon request .j Footnotes to i 50.55a; pursuant to i 50.55atax2H11).

                                                                   ' For purposes of this regulation, the pro-
                ' (Reserved]
                ' Components which are connected to the posed IEEE 219 became "in effect" on                           -

August 30.1968, and the revised tasue IEEE reactor coolant system and are part of the 279-1971 became "in effect on June 3.1971. l j reactor coolant pressure boundary defined . Copies may be obtained from the Institute in 150.2(v) need not meet these require. of Electrical and Electronics Engineers, ments, provided.' United Engineering Center. 345 East 47th : la) In the event of postulated failure of ' Street. New York NY 10017. A copy is avall-the component dunns normal reactor oper- able for . inspection at the Commission's ation, the reactor can be shut down and Public Document Room.1717 E Street NW cooled down in an orderly manner, assuming - Washington. D.C. makeup la provided by the reactor coolant

Where an application for a construction makeup system only, or pennit is submitted in four parts pursuant (b) The component is or can be isolated to the provisions of I 2.101(n.1) and Subpart from the reactor coolsot system by two F of Part 2 of this chapter, "the formal- -)

valves (both closed, both open, or one closed docket date of the application for a con- - l and the other open). Each open valve must struction permit" for purposes of this sec. be capable of automatic actuation and, as- tion shall be the date of docketing of the in. suming the other valve is open, its closure formation required by 12.101(a-1)(2) or,(3), 'I time must be such that, in the event of pos- whicheve' ir Ster. -

                                                                                              -                                   J tulated failure of the. component during- I 50.56 - Conversion of construction permit                                i normal reactor operation, each valve re.

mains operable and the reactor can be shut to license: or amendment of license. q down and cooled down in an orderly Upon completion of the construction manner, assuming makeup is provided by or alteration of a facility, in compli-the reactor coolant makeup system only. ance with the terms and conditions of 8 Copies may be obtained from the Ameri- the* construction permit and subject to can Society of Mechanical Enstneers, United Engineering Center, 345 East 47th ~any necessary testing of the facility

St., New York NY 10017. Copies are avails. for health or safety. purposes, the ble for inspection at the Comm halon's Commission will, in the absence of Public Document Room 1717 H St. NWa good cause shown to the contrary Washington, D.C. issue a license of the class for which

                  'USAS and ASME Code addenda lasued the construction permit was issued or ered          in effe " or e oc ve f an appropriate amendment of the li-cense, as the case may be, months after their date of issuance and after they are incorporated by reference in (Sec.185. 64 Stat. 955; 42 U.S.C. 2235) paragraph (b) of this section. Addenda to (21 FR 355. Jan.19.1956, as amended at 35 the ASME Code lasued after the Summer FR 11461, July 17,1970)                                                       ,

1977 Addenda are considered to be "La ! effect" or " effective" after the date of publi- - 8 50.57 1ssuance of operating fleense.8 ' cation of the addenda and after they are in. (a) Pursuant to i 50.56, an operating . corporated by reference in paragraph (b) of license may be lasued by the Commis. ( this section. 2

                   'For ASME Code Editions and Addenda sion, up to the full term authorized by                                    l lasued prior to the Winter 1977 Addenda, i 50.51, upon finding that:                                               J the Code Edition and Addenda applicable to .           (1) Construction of the facility has the component is governed by the order or been substantially completed, in con.                                 -

contract date for the component, not the formity with the construction permit contract date for the nuclear energy system. and the application as amended, the For the Winter 1977 addenda and subse, quent editions and addenda the method for provisions of the Act, and the rules 4 i determining the appI! cable Code editions and regulations of the Cotnm(ssion; and addenda is contained in Paragraph NCA and 1140 of Section III of the ASME Code. 'The Commiazion may issue a provisional ;

                    'ASME Code cases which have been de- operating license pursuant to the regula.                                 a termined suitable for use by the Commis, tions in this part in effect on March 30.                                    l sion staff are listed in NRC- Regulatory            1970, for any facility for which a notice of                    q Oulde 1.44. " Code Case Acceptability-a hearing on an appilcation for a provisional                                s ASME Section m Design and Fabrication              operating license or a notice of proposed is.                     l and NRC Regulatory Guide 1.85. " Code               suance of a provisional operating ilcense has Case Acceptability-ASME Section m Ms.              been published on or before that date.

terials." The use of other Code cases may be 437 l 06/01/84 134 Revision A !

l 95053 Title 10-Energy ; (2) The facility will operate in con- this section as to which there is a con- W troversy, in the form of an initial deci. formity wit h the application as amended, the provisions of the Act, sion with respect to the contested ae- ; and the rules and regulations of the livity sought to be authorized. The Di- ' Commission: and rector of Nuclear Reactor Regulation (3) There is reasonable assurance (i) will make findings on all other matters

                .that the activitics authorized by the         specified in paragraph (a) of this sec-operating lleense can be conducted            tion. If no party opposes the motion.                                                       >

without endangering the health and the presiding officer will issue an j safety of the public, and (11) that such order pursuant to 12.730(e) of this I activitics will be conducted in compli- chapter, authorizing the Director of l ance with the regulations in this chap- Nuclear Reactor Regulation to make ter; and appropriate findings on the matters * (4) The applicant is technically and speelfled in paragraph (a) of this sec-financially quallfled to engage in the tion and to issue a license for the re- . activitics authorized by the operating quested operation. l license in accordance with the regula- l tions in this chapter. However, no (35 FR 5318, Mar. 31,1970, as amended at finding of financial qualifications is $ 6 "' necessary for an electric utility appll* 2 ^d 151I k y 1972. FR l 8790, Mar. 3.1975: 47 FR 13755, Mar. 31, ! cant for an operating license for a pro

19821 (

duction or utilization facility of the type described in i 50.21(b) or 150.22. 8 50.58 Hearings and report of the Advino-(5) The applicable provisions of Part ry Committee on Reactor Safeguards. 1 0 of this chapter have been satisfied; (a) Each application for a construc-(S) The.lssuance of the license will tion permit or an operating license for l not be inimical to the common defense a facility which is of a type descritred and security or to the health and in i 50.21(b) or 150.22, or for a testing safety of the public. facility, shall be referred to the Advi- , (b) Each operating license will in- sory Committee on Reactor Safe- j l clude appropriate provisions with re. guards for a review and report. An ap- l spect to any uncompleted items of plication for an amendment to such a l construction and such limitations or construction permit or operating II. l conditions as are required to assure cense may be referred to the Advisory i that operation during the period of Committee on Reactor Safeguards for l the completion of such items will not review and report Any report shall be l endanger public health and safety, made part of the record of the applica-(c) An applicant may, in a case tion and available to the public, except where a hearing is held in connection to the extent that security classifica. with a pending proceeding under this tlon prevents disclosure. section make a motion in writing, pur. (b) The Commission will hold a suant to this paragraph (c), for an op- hearing after at least 30 days notice crating license authorizing low power and publication once in the FEDERAL testing (operation at not more than 1 REctsTER on each application for a percent of full power for the purpose construction permit for a production of testing the facility), and further op. or utilization facility which is of a - erations short of full power operation. type described in 160.21(b) or i 50.22 Action on such a motion by the presid. or which is a testing facility 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 proceedings, including the right of ing of a public hearing and an applica-any party to be heard to the extent tion is made for an operating license that his contentions are relevant to or for an amendment to a construction l the activity to be authorized. Prior to permit or operating license, the Com-taking any action on such a motion mission may hold a hearing after at which any party opposes, the presid- least 30 days notice and publication ing officer shall make findings on the once in the ProtaAt. RzarsTER or, in matters specified in paragraph (a) of the absence of a request therefor by 438 06/01/84 135 Revision A

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                                                                                                                                                       .)

Chapter I-Nuclear Regulatory Commission { 50,70 q q any person whose interest may be af. ments carried out pursuant to parn- 1 fected, may issue an operating license graph (a) of this section. These rec- J or an amendment to a construction ords shall include a written safety j permit or operating license without a evaluation which provides the bases hearing, upon 30 days notice and pub- for the determination that the change. ] llcation once in the ProtnA1, RectsTrn test or experiment does not involve an of its wern to do so. If the Commis- unreviewed safety question.L The 11-slon fode thn ro significant hazards censee shall furnish to the appropriate ) considerd% ta pNsented by an appil- NRC Regional Office shown in Appen. cation for an .e* ndment to a cor- dix D of Part 20 of this chapter with a struction permit or operating license, copy to the Director of Inspection and it may alspense with such notice and Enforcement, U.S. Nuclear Regulatory publication and may lasue the amend- Commission. Washingto 1, D.C. 20555, 1 ment. annually or at such shorter htervals 4 (27 FR 12186 Dec. 8,1962, as amended at 33 as may be specified in the 'llcense, a FR 8590. June 12.1968; 35 FR 11461. July report containing a brief description 17,1970; 39 FR 10555, Mar. 21,1974) of such changes, tests, and experi-ments, including a summary of the 9 60.59 Changes, tents and experiments. safety evaluation of each. Any report ! (a)(1) The holder of a IMense submitted by a licensee pursuant to authorizing operation of a production this paragraph will be made a part of l I or utilization facility may (1) make the public record of the licensing pro- j l changes in the facility as described in I ceeding. In addition to a signed origi. the safety analysis report (11) make nal, 39 copies of each report of changes in the procedures as described changes in a facility of the type de-in the safety analysis report, and (!!!) scribed in 150.21(b) or i 50.22 or a 'l

                                                                                                                                                        .I conduct tests or experiments not de-        testing facility, and 12 copies of each scribed in the safety analysis report, report of changes in any other facility,                                                            1 without prior Commission approval, 'shall be filed. The records of :hanges unless the proposed change, test or ex*                                                                                                     {

in the facility shall be maintained J l periment involves a change in the until the date of termination of the 11 I technical specifications incorporated cense, and records of changes in proce, j in the license or an unreviewed safety dures and records of tests and experl* question, (2) A proposed change, test, or ex. ments shall be maintained for a period of five years. .' periment shall be deemed to involve an unreviewed safety question (1) If (c) The holder of a lic Izing operation of a prodt'ense ictionauthor. or uti-the probability of occurrence or the !!zation facility who desires (li a consequences of an accident or mal

function of equipment important to change in technical specifications or safety previously evaluated in the (2) to make a change in the facility or safety analysis report may be in. the procedures described in the safety I analysis report or to conduct tests or creased; or (11) If a possibility for an experiments not described in the accident or malfunction of a different s&fety analysis report, which involve type than any evaluated previously in an unreviewed safety question or a the safety analysis report may be cre' sted; or (111) If the margin of safety as change in technical specifications. -

defined in the basis for any technical shall submit an application for amend. specification is reduced. ment of his !! cense pursuant to 4 50.90. (b) The licensee shall maintain rec- (39 FR 10555, Mar. 21,1974, as amended at ords of changes in the facility and of 41 FR 16446, Apr 19.1976; 41 FR 18302, changes in procedures made pursuant May 3,1916; 42 FR 20139. Apr,18.19771 to this section, to the extent that such changes constitute changes in the is. Instretions, Reconos, Rtronts, cility as described in the safety analy. NoTIrtcAT1ons als report or constitute changes in pro-cedures as described in the safety 4 50,M inspections. 1 analysis report The licensee shall also (a) Each licensee and each holder of ' maintain records of tests and experl. a construction permit shall permit in-439 06/01/84 136 Revision A

                                                                                                 ,      ..a. . _ . - - . - . - - - _ - _ - - - . -

-{

s l e l l

l 9 50.90 Title 10-Energy with the regulations in this chapter Revocmon, Busernston Montrva-and will not be inimical to the T1oN. AMENouzwT or LICENSES ' arD . common defense and security or to the CONSTRUCTION PERMITS, ExtacENey health and safety of the public. - OrzRATroNs av THE Commisstow ' (b) If the application demonstrates that the dismantling of the facility 8 50.100 Revocation. eunpension. rnadifice and disposal of the component parts tion of licenses and construction per 4 will be performed in accordance with " ' l" I* f "- l the regulations in this chapter and A license or construction permit may will not be Inimical to the common de- be revoked, suspended. or modified, in l fense and security or to the health whole or in part, for any material false l ' and safety of the public, and after ' statement in the application for ll- { notice to interested persons. the Com- cense or in the supplemental or other ' mission may issue an order author- statement of fact required of the ap-izing 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, inspection, or other means, conditions specified in the order, which would warrant the Commission (26 FR 9546. Oct.10.1961, as amended at 32 to refuse to grant a license on an origi. 1 FR 3090. Feb. 21,19671 nal application (other than those re ' ' lating to 15 50.51, 50A2(a), and AurwnutwT or LictNsr oR ConsTauc. 50A3(t,) of this part); or for failure to Tion PEam2T AT Rreurs7 or Ho!. Den construct or operate a facility in ac-cordance with the terms of the con. 9 50.90 Appliention for amendment of IL struction permit or license, provided cease.er eenettwtion permit. that failure to rnake timely completion of the proposed construction or alter. Whenever a holder of a license or construction permit desires to amend ation of a facility under a construction the license or permit, application for permit shall be governed by the provi- +l an amendment shall be filed with the sions of I 50.55(b); or for violation of, I Commission, fully describing,, _ the or failure to observe, any of the terms and provisions of the act, regulations, i changes desired, and foIIoking as far license, permit, or order of the Com- i as applicable the form prescribed for 'mi .str n' 6rlglnifTpplications. '- e j, 8 50.91 Innuance of amendment, 8 50.101 Retaking possession of special - ' nuclear material. In determining whether an amend

Upon tevocation .of a license, the ]

ment to a license or construction Commission may immediately cause i i permit will be issued to the applicant the retaking of possession of all spe-the Commission will be guided by the clal nuclear material held by the 11

considerations which govern the issu; censee, t

ance of inttial licenses or construction permits to the extent applicable and 121 FR 355, Jan.19,1956, ma amended at 40 appropriate. If the application in. FR 8M0, Mar. 3. H751 - volves the material alteration of a 11* 8 60.102 Commission order for operation censed facility, a construction permit will be issued prior to the issuan:e of after revocation. I 1 the amendment to the license. If the Whenever the Commission finds - amendment involves a significant haz, that the public convenience and neces. ( ards consideration, the Commission sity, or the Department finds that the will give notice of its proposed action production program of the Depart, pursuant to i 2.105 of this chapter ment requires continued operation of before acting thereon. The notice will a production or utilization facility, the be issued as soon as practicable after license for which has been revoked, the app!! cation has been docketed, the Commission may, after consulta-tion with the approprhte federal or (39 FR 13258. Apr.12,19741 state regulatory agency having jurts-444 t I 06/01/84 137 Revision A

Chanter 1-Nuclear Regulatory Commission Part 50, App. A , diction, order that possession be taken compliance with the rules, regulations, of such fnellity and that it be operated or orders of the Commission. for a period of time as, in the judg- (c) The Commission may at any time ment of the Commission, the public require a holder of a construction convenience and necessity or the pro- permit or a license to submit such in-duction program of the Department formation concerning the addllion or may require, or until a license for op- proposed addition, the elimination or eration of the facility shall become ef- proposed elimination, or the modifica-fective. Just compensation shall be tion or proposed modification of struc-paid for the use of the facility. tures systems or components of a fa-cility as it deems appropriate. (40 FR 8790. Mar. 3.19751 0 * "" 8 50.103 Suspension and operation in war or national emergency. ENFORCEMENT (a) Whenever Congress declares that a state of war or national emergency a 56.110 Violationn. exists, the Commission, if it finds it An injunction or other court order necessary to the common defense and may be obtained prohibiting any viola-security, may, .tlon of any provision of the Atomic (1) Suspend any licent.e it has issued. Energy Act of 1954, as amended, or (2) Cause the recapture of special Title II of the Energy Reorganization nuclear material. Act of 1974 or any regulation or order (3) Order the operation of any 11- !ssued thereunder. A court order may censed facility. be obtained for the payment of a civil (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 material or to operate the facill- 53, 57, 62, 63, 81, 82, 101, 103, 104. 107, ty, or 109 of the Act, or section 206 of the (b) Just compensation shall be paid Energy Reorganization Act of 1974 or for any damages caused by recat ture any rule, regulation, or order issued ' of special nuclear material or by c.7er- thereunder, or any term, condition. or ation of any facility, pursuant to Ols limitation of any license issued there-section. under, or for any violation for which a (Sec.108. 68 Stat. 939, as amended: 42 license may be revoked under section U.S.C. 2138) 188 of the Act. Any person who will-(21 FR 355. Jan.19.1956, as amended at 35 fully violates any provision of the Act FR 11416. July 17,1970; 40 FR 8790, Mar. 3 or any regulation or order issued 19751 thereunder inay be guilty of a crime nd, upon Conviction, may be punished BACMFITTING by fine or imprisonment or both, as R 50.109 Backfitting. provided by law. (a) The Commission may, in accord. (40 PR 879o. Mar. 3.1975, as amended at 42 ance with the procedures specified in Fl} 25721. May '19,19771 this chapter, require the backfitting of a facility if it finds that such action APPENDICES will provide substantial, additional - protection which is required for the

   .           public health and safety or the               . ArrENoix A-GENERAt. DEstoN common defense and security. As used           CRITERIA FoR Nuct.zAn PowEn Pt. ANTS in this section, "backfitting" of a pro-duction or utilization facility means                       M ' 8/ C*""'8 the addition, elimination or modifica-                       turnooverson tion of structures, systems or compo.

nents of the facility after the con. otrintmns struction permit has been Issued. Nuclear Power Unit. (b) Nothing in this section shall be Loss of coolant Accidents. , deemed to relleve a holder of a con- slasie Failure. I struction permit or a license from Anticipated Operational Occurrences. l 445 06/01/84 130 Revision A

    ,                                               ((J'                                                                    UNITED STATES                                       d
  ..           8                       i                             NUCLEAR REGULATORY COMMISSION h                       :$                                                                                  WASHINJTON, D. C. 20555 4

4 %*****/ May 17, 1985 CHAIRMAN T h e H o n o r a b l e E d w a r d J . M a r k eM) SLO 90ttslith' ROCM , Subcommittee on Energy Conservation and Power Committee on Energy andReprese Commere atNekg-NM5 United Stater House of ! Washington, DC 20515

Dear Mr. Chairman:

1 Recently, Mr. Licciardo, an NRC staff member, met with'me under NRC's Open Door Policy regarding the Commission's letter to you dated December 20, 1984 on the subject of erroneous McGuire Technical Specifications. He felt that the December 20, 1984 letter mischaracterized his involvement.in the review of the McGuire Technical Specifications and that his actions.were inaccurately cited as the main cause for delay in resolving his differing professional opinion (DPO) on these same specifications. This letter.is intended to-correct any mischaracterizations or misrepresentations regarding Mr. Licciardo in our December 20 letter. Our December 20 letter should not have inferred that Mr. Licciardo introduced unnecessary delays nor that the detailed attention provided during the' staff's review resulted in unwarranted or avoidable delays. The problem is complex and, I as such', is not subject to singling'out one cause.of delay. l Due to the sheer magnitude of his concerns, over 300.in all, it l took a significant amount of time for Mr. Licciardo to provide the required bases for each item'. Likewise, a significant and ; , lengthy staff effort was necessary to evaluate each item. Based on my conversation with Mr. Licciardo and his subsequent discussions with my personal staff, I believe the pace of the , staff's review is acceptable to Mr. Licciardo. The staff found j in February 1984 that none of.the McGuire concerns presented an

imminent public health or safety problem. Given this finding ,

I and the increased attention afforded by the staff to this ! matter, I believe that the McGuire Technical Specification j evaluation is proceeding at a satisfactory pace, j l Mr. Licciardo also indicatea that the December 20, 1984 letter to you mischaracterized the present state of the McGuire i Technical Specifications. However, I have not been able to i confirm Mr. Licciardo's claim. As I noted above, the staff l made an initial finding that there was no imc11nent safety j s w---_-_-_-_-__-_-_.-______________--_-_.-_-.___-- _ _ _ - - . _ . . - _ - - - - - - _ _ _ _ . _ _ _ . _ - _ . . _ _ - - _ _ - - _ _ . . _ -_- _- - . . - - - _ _ _ - _ - _ - _ _ _ _ _ _ - - - -

' ' * ~ I: 2 problem with the Technical Specifications. The.380 items identified by Mr. Licciardo were evaluated by e' team of reactor 1 systems technical managers. That'_ team concluded that 160 of I the items did not warrant further attention either because: (1) Mr. Licciardo's assessment-of the issue was. incorrect, or (2) the management team (all of whom were experienced reactor systems re' viewers) could not understand Mr. Licciardo's description of the issue. 3

                 'The management team concluded that the remaining 220 did              ,

warrant additional NRC evaluation. The present schedule calls j for completion of the staff evaluation and categorization of l those 220 items by late spring of this year. Upon completion 1 of this categorization a letter will be forwarded to the-licensee requesting his response to plant specific issues .; within three months. The remaining issues of the 220 items ! which are generic in nature will be handled as part of our l generic issues program with a target. date for final resolution by the end of this year. This letter and all-subsequent letters, will be a matter for the public record, and, as such, will be docketed. If any information becomes available which causes us to reconsider the staff's initial finding, the f schedule will be accelerated. I appreciate Mr. Licciardo's sincerity and conscientiousness in bringing his concerns to my attention. I trust that this letter will further clear the air on his involvement in the schedule of resolving the concerns arising from his Differing Professional Opinion. Sincerely, qg , r7 l

                                                    /(LL       Y;       (bt k m .-

Nunzios J. Palladino cc: Rep. Carlos Moorhead

Enclosure 3 6; C-JL A T OR Y INFORMATION DISTRIBUTION SYSTEM (R IDS) aCCEEE:ON ND R . E t :5190394 DOC.DATE: 86/06/10 NOT ARI Z ED: NO DOCKET e FACI- 50-359 William B. McGuire Nuclear Station, Unit 3, Duk e Pove 05000369 D 27 W.11:am P. McGuire Nucle.ar Etation,' Unit 2

                         .                                                      Duke Pcce        05000370 AUT- WA'!D                AUTHOR AFFILIATION 1UCKER.M I<               DuLe P o ure r Co.

3ECI: NAM 7 f(F C I P I E NT AFFILIATION . LENTO.% d h. Off3cc of Nuclear Reactor Regulation, Direc tor (p os t 851125 YOUNC-D OOD, B. J FW3 Project Directorate 4 EVDvE: Retponec to 850709 Itr te plant-specific concerns from review o# m:d-Jan 1993 Tech Specs. Tech Spec 4< FSAR revs vill be pursued upon NRC concurrence v/ positions. DISTR:buTION CODE: AOO1D COPIES RECEIVED: LTR d ENCL SIZE: 111 LE. On Submi t t c1: General Distribution l ND1 E 5 ' I 1 RECIPIENT COPIES RECIPIENT COP]ES JD COD!!/NAME LTTR ENCL ID CODE /NAME L1TR ENCL PW3-A Eb 1 1 PWP-A EICSD 2 2 PUN A F03 J 1 PWR-A PD4 LA J O PW - ' AD4 PD 01 D 5 PWR-A PD4 OJ 5 5 q 1 1 PWR-A PSD 1 1 i im,< - A R $b ; 1 INT 2RNAL ADM/LFND 1 O ELD /HDS4 1 O I NHR/DHrT/T6CB 1 1 NRR/ORAS 1 0 \ FtF G F IL E OA 1 1 RGN2 1 1 l E X TER NA . ECbG DRUSHE,5 1 1 LPDR 03 1 1 l MRC PDR 07 1 1 NSIC 05 1 1 l 1 l l t I i

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l h i l i TCT L C':!G CF CCr.I?S REQUIP;D. LTTR 2E ENCL 24 i l

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DUKE POWER GO> IPA >'t P.O. BOX 33189 ! CILARLOrTE. N.C. 28242

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j June 10, 1986 1

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                                  .,....<.~,...;4,    Director                           4 e . c .. .uc eo       actor Regulation                                                      I U.S. Nuclear Regulatory Comnission                                                            I Washington, D.C. 20555 ATTENTION:       3.J. Youngblood, Director PWR Project Directorate #4

Subject:

McGuire Nuclear Station Docket Nos. 50-369 and 50-370 NRC DP0 Concerns on McGuire Technical Specifications

Dear Mr. Denton:

f Mr. T.M. Novak's (NRC/0NRR) July 9,1985 letter to Mr. B.B. Tucker L (DPC) indicated that a review of the McGuire Unit 1 and 2 Technical Speci- { l fications was being conducted in response to concerns raised by a member of the NRC staff in a differing professional opinion (DPO) resulting from a 3 review of the proof and review copy of the McGuire Unit 1/2 combined Tech- ! nical Specifications which existed in mid-January 1983. Duke Power Company's co==ents were requested on certain plant-specific concerns contained in the DPC (other concerns contained in the DP0 were either being considered by the ' NRC for generic resolution, had been closed by NRC internal review, or were still under review). I Attached is Duke Power Company's response to these concerns. This response is limited to the specified plant-specific concerns and doss not address any f generic aspects oL these specified concerns. Note that the response has i potential plant-specific impacts on the station's Technical Specifications (e.g. question 4a&b, and 4c). nos. 6a, 7d (and 71, 7k), and 7n) and FSAR (e.g. questions Duke vill pursue appropriate plant-specific Technical Speci- , i fication and FSAR revisions following NRC concurrence with the positions contained herein. The Westinghouse Standard Technical Specification issues identified in this response should be resolved on a generic bc. sis (note that Westinghouse reviev/ input was utilized in the development of this response), l j Note also that generic Technical Specification improvement effortr currently i underway by industry (e.g. AIF, WOG, B&WOG) and NRC (TSIP) may impact the DPO's concerns and the resolutions proposed by this response. j j As indicated above, the NRC is requested to approve this response prior to Duke proceeding with the appropriate Technical Specification change submit- l tais and inclusion of the informaitou in a future FSAR update. Should there f

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_ _ _ - _ _ _ _ _ _ . I

a . .I I 4 J j Mr. Earold R. Denton, Director-June 10,19E 6 Page 2

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be any questions regarding this matter or if additional information is required, please advise. ; Very truly yours' ,

y k1 / 9' Spe.;- Ea1 B. Tucker ' i PBS/j g= i Attachment l xc: Dr. J. Nelson Grace, Regional Administrator U.S. Nuclear Regulatory Commission - Region II 101 Marietta Street, tiW, Suite 2900 l Atlanta, Georgia 30323 Mr. Darl P,od i Division of Licensing Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Washington, D.C. 20555 l Mr. W.T. Orders

Senior Resident Inspector McGuire Nuclear Station 4 Ms. L.L. Williams, Manager ESSD Projects, Mid-South Area 'j Westinghouse Electric Corp.

HNC West Tower P.O. Box 355 Pittsburgh, PA 15230 l l i l 4 1 L - - - - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - . - - - - _ - - - - _ __

I i Duke Power Co=pany . l McGuire Nuclear Station Respense to NPC DP0 Concerns

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J (Questict l' TAELI f. 2- 1 .e These have been checked against reference 18, Westinghouse (W) RPS/ESFAS-Set Point' Methodelegy, Table 3-4 and NOTE FOR TABLE 3-4 on page 3-13, which is described as applicable to McGuirefUnit 1, 50-369. At this date, the' assumption , has been made that this information also applies to McGuire Unit 2,-Docket No. 50-370. Please docket this fact or otherwise provide the alternate information Respecse: ,The data contained in Reference 18 bas been confirmed to be valid i for botb.McGuire Unit 1 and Unit 2. The instrumentation hardware: (racks, transmitters) are tbt same'for both Units 1 and'2. While the SteamLGenerators are different (D-2 for Unit 1 and D-3 for Unit 2), there are no differences-in the Safety Analysis. values.s :There-- f ore it can be concluded that the Setpoint Study performed for Unit 1 is applicable, in it's entirety, to Unit 2. The safety analysis 'l j

                      ' performed is valid for both units and use the same. equipment /          ,

instrumentation resulting in uncertainty values being valid for both ! units. l (Question la) T AB LE 2 . 2 - 1. Item 3 l Will a time constant of >2 seconds result in a slower respcase time, which is. less conservative. i I Response: The dynamic response of the High Positive Rate trip function is i similar to the rate / lag function associated with the AT trips. The i responses of the various dynamic functions are demonstrated in Appendix A of VCAP-8745.(Design Bases for the Thermal Overpower AT , l and Thermal Overtemperature AT Trip Functions). As may be seen in ! the above mentioned figures, an increased time constant results-in faster response and thus results-in a shorter time from initiation of .] j transien:. to reactor trip. Therefore, the >2 seconds Tech Spec j i, requirement for the time constant is conservative. ' (Question Ib) TABLE 2.2-1, Item 4 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 'i 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 Licensee shall evaluate and propose.

w a-Respecse: The dynamie response of the High Negative Rate trip function is similar to the rate / lag function associated with the AT trips. -The responses of the various dynamic functions are demonstrated in Appendix A of WCAP-8745 (Design Bases for the Thermal Overpower AT and Thermal Over temperature AT Trip Functions). As may be seen in the above mentioned figures, an increased time constant results in faster response and thus results in 'a shorter time from initiation of transient to reactor trip. Therefore, the >2 seconds Tech Spec requirement for the time constant is conservative. The Revision 7 FSAR analysis referred to in this inquiry was performed prior to the NRC review and approval of WCAP 10297-P-A (Dropped Rod Methodology For Negative Flux Rate Plants). The methodology used prior to WCAP-10297-P-A did not involve an actual 1 i

                                 . determination of the negative flux rate setpoint and/or determination of the maximum dropped rod (s) worths which might not-l result in a reactor trip. The statemect in the FSAR (RCCA group         j results in reactivity insertion of s-1200 pcm which results in a         I reactor trip within s 2.5 seconds) was meant only to offer support for the DNB analysis performed at lower rod worths but did not actually demonstrate the adequacy of the negative flux rate setpoint.

l Upon determination of possible nonconservatisms in the analytical methodology, Westinghouse developed the dropped rod methodology

outlined in WCAP-10297-P-A. The revised methodology links the ! assumptions regarding the negative flux rate setpoint, rod worths and locations, control system behavior, and other factors which iniluence plant behavior following a dropped rod (s) event. The l setpoint thus becomes an integral part of the safety analysis and ths table in reference 18 is revised to show a safety analysis limit {' of 6.9'. RTP. The adjustments made to account for various uncertainties results in an STS Trip Setpoint of 5.0% RTP and an STS Allowable Value of 5.5% RTP. Details regarding the revised methodology and basis for the.setpoint may be found in WCAP-10297-P-A. (Question Ic) TABLE 2.2-1, Item 9 The specified Trip Setpoint & Allowable values agree with those provided under setpoint methodology in reference 18. A disparity does exist between the related SAFETY ANALYSIS LIMITS given as used in Safety Analysis, i.e., 1845 psig in SETPOINT METHODOLOGY / reference 18, Table 3-4, column 12 and the FSAR value for the same analysis in reference 7, Table 15.2.3-1 as 1835 psig. The Licensee shall identify the correct value. [ Note also disparity with reference 7, " Analysis of Inadvertent Operation of ECCS During Iower 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 respect to the other values used above.) The Licensee shall review and clarify.

1 i 1 l

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hesponse: The analysis of the inadvertent operation of ECCS-during power f operation had assumed a low pressure setpoint of 1800 psia while ) other analyses assumed a setpoint of 1835 psig. The reference 18 ' value for the safety analysis limit was in error but was conservative and since margin exists between implemented and required setpoints,- the conservatism did not impact the trip " setpoint and allowable values. , , , The transient analyses have been reanalyzed as a result of the transition to optimized fuel assembly design. The revised analyses assumed a safety analysis limit of 1850 psia (1835 psig) for all transients. (Question Ic) TAB LE 2. 2- 1. Ite 13 Reference 18, page 3-13, Note 12 describes the S.afety Analysis Limit for this J item as a value in Table 2.2-1 of the W STS plus' 10%. For conservatism, should the Safety Analysis Limit be the y STS value less 10%; is this necessarily conservative for all Licensing Basis occurrences? Response: The analysis in effect at the time this question was posed is no 1 longer applicable. At present the bounding analysis for the steam ! generator lo-lo level-is the feedbreak analysis. This analysis is done assuming the system starts at full power. In this aculysis the safety analysis limit is 23% of narrow range span. As is indicated in the technical specifications this corresponds to a nominal trip setpoint of 40% narrow range span at 100% RATED ' Ti[ERMAL POWER. (Question le) TABLI 2. 2-1, Item 18b Accidental Depressurization of the main steam system is from zero load. It is unclear from reference 5 Table 7.2.1-4, (page 5 of 5) if for this event, l 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 j initiates SI, or from SI initiated by other initiators. The Licensee shall j clarify, and hence its validity with respect to the absence of the signal i caused by P-7. Response: Protection against accidental depressurization of the main steam I system is provided by the overpower reactor trips (neutron flux and AT) and by the reacter trip which results from the receipt of the safety injection (SI) signal. The tafety injection signal is actuated by low steamline pressure, low pressurizer pressure, or j l l l l L J

w l

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l l l 1 tish containr:ent pressure. The analysis performed results in SI initiation on low pressurizer pressure and reactor trip will. I either occur concurrently due to the trip on SI actuation or will occur prior'to SI on the overpower trips. The main steam , depressurization analyzed in the FSAP is initiated from hot shutdown conditions at time zero (i.e. reactor tripped) since this represents the most conservative initial condition. Thus no q explicit assumption is made regarding the cause of reactor trip for the FSAR analysis. As noted in the FSAR and above, should the reactor be just critical or operating at power a reactor trip wom occur on the overpower trips or from an SI actuation. In either case, no credit is taken for the reactor trip on pressurizer pressure when reactor power is below the P-7 interlock. (Question 2) T.S. Page 3/4 1-6 The existing minimum temperature for criticality (In MODES 1 and 2) is given as 551*F. Please advise why this value is less than the programmed set point minimum value of 557'T in reference 20, Fig. 5.3.3-1. Accident evaluations ; l f or events f rom zero power are predicated upon this set point -of 557'F, and ) acy variation therefrom in either direction would be unacceptable, j Response: TSAR Figure 5.3.3-1 gives the normal relationship between reactor coclant system temperature and power. The hot zero power temperature employed at McGuire and used in the safety analysis is $57'F. The minimum temperature for criticality is determined such that the moderator temperature coefficient is within its ; analyzed temperature range,'the trip ~ instrumentation is within its operating range, the pressurizer is capable of being in an operable I status with a steam bubble, and the reactor vessel is above its minimum RT ! temperature. The minimum temperature for criticality limitintbTMcGuire Technical. Specifications is 551'F. j The difference between the HZP temperature and minimum temperature for criticality limit is required in order to allow for measurement l of the moderator temperature coefficient. Since the moderator coefficient is confirmed to be within safety analysis assumptions at ~ conditions of~approximately 551*F - 557'F, the only input parameter to the safety analysis of concern is the initial temperature. The change in initial conditions from 557'T to 551'T for transients occurring at HZP would have a negligible impact on results and would be a less representative input since the majority of time spent at RZP conditions includes temperatures of $557'F. As noted, the accidents analyzed at hot zero power (HZP) assume an RCS temperature of 557 'F. The FSAR notes that use of a higher initial system temperature yields a large fuel-water heat transfer coefficient, larger specific heats, and a less negative (smaller absolute magnitude) Doppler feedback effect for fast reactivity addition transients like the RCCA Bank Withdrawal from Subcritical and HZP Rod Ejection events. The reduced feedback results in a l

I j

l l 1 higher neutron flux peak. For a Steamline Break event, starting

        ,                     from a aigher initial RCS temperature results in a greater increase t

in coclant density from the cooldown. More reactivity is added due to the positive moderator density coefficient and a higher return to power results when compared with.the case of a lower initial RCS temperature. Based on these considerations, a higher initial RCS l temperature is conservative for the analysis of events from power. Tne statement that any variation in HZP temperature is unacceptable is also not consistent with the general conservative philosophy j used to evaluate nuclear plant safety since only limited analyses l are performed to demonstrate adequate safeguards for a range of l plant conditions, i l l (Question 3) T A3 LE 3 . 3 - 1, Item 6e During shutdown in MODES 3, 4 and 5, with reactor trip system breakers open, Source Range, Neutron Flux, channel operability requirements specify only one chaanel operable, and if this same channel is being used to meet the boron dilution alarm requirements of proposed T.S. Page 3/4 .1-13 (a), then it is not l in accordance with the Boron Dilution Requirements of the FSAR for which at - least 2 operable chaanels would be required; reference 8, page Q 212-24. Item 212.56. The Licecsee shall evaluate and propose. Currently, this appears- i l non-conservative. I Response: A review cf FSAR Section 15.4.6 (Boron Dilution Accident) does not 2ndicate the number of Source Rsage Channels required operable; however, These channels are mentioned for Refueling (MODE 6) and l start up (MODE 2) Dilution Accidents. For these cases, two ! chaanels 3ce required per Tech. Specs. Additionally, MODES 3,4, I and 5 are not addressed by this FSAR Section. Boron Dilution analyses during MODES 3,4, and 5 are not part of the McGuire plant licensing basis. As such, any channel operability requirements would not be based on the FSAR analysis. Generic Letter 85-05 dated January 31, 1985 informed licensees of the Staff position resulting from the evaluation of Generic Issue 22 " Inadvertent Boron Dilution Events". The Staff concluded that the consequences of such events are not severe enough to jeopardize the health and safety of the public. Furthermore,' while NRC stated that it would "not require operating plant backfits for boron dilution events at this time, the staff would regard an unmitigated boren dilution event as a serious breakdown in the licensee's ability to control its plant, and strongly urges each licensee to assure itself that adequate protection against boron dilution events exists in its plants". McGuire personnel believe that adequate protection against boron dilution events exists and that no changes to technical specifications are warranted in this instance. l 1 l l l l l

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l (Questien La and Ab) i TAELE 3.3-2. Items 9 & 10 - The T.S. spec:fies a response time of 5 2.0 secs. Reference 7, Table ! j 15.1.3-1 provides a time delay of 2.0 sees for these events,which conflicts ; w:th a value of 1.0 secs in Reference 5, page 7.2-14, rev. 42, item 1(e). ! The Licensee shall clarify. ' Response: The Technical Specification limit of f 2.0 seconds for the time delay of pressurizer pressure trip functions (low and high) is based upon th. FSAR Chapter 15 transient analysis which assumed a l delay of 2.0 seconds. The values for trip response times in ! chapter 7 are " typical maximum allowable time. delays" and are not ; necessarily the same as the McGuire specific assumptions. For the ! sake of clarity, the values provided in chapter 7 will be revised to agree with Chapter 15 and Technical Specifications in a future , l FSAR update. 1 (Question 4c) i TABLE 3.3-2, Itet 17 I The proposed T.S. states that the response time, requirement is NA (Not Applicable). This is incorrect since a separate Reactor Trip is an essential j part of all ESTAS functions during which safety injection is initiated. The required :nformation is in fact supplied in T.S. Page 3/4 3-30 Table 3.3-5, under Jb, 4b. the already revised headings proposed above, Reference Items 11, 2b, ' This table, under response time, should replace the description as recommended above and alongside each, reference the entry in T.S. Table 3.3-5. The response given in the Technical Specifications (except for manual actuation of SI) are quoted as 52 secs. No docketed information is available on what values were used in accident analysis, and particularly for MSLB, SBLOCA and LOCA events. The licensee should provide this information and confirm its conservatism against the T.S. value, e.g. reference 5, Table 7.2.1-4 (5 of 5) and related Note e on the page entitled " Notes for Table 7.2.1-4" confirms that Pressurizer Low Pressure - Low Level is the first out trip'of

Safety Injection for the event of " Accidental Depressurization of the Main Steam System." The licensee shall explain this terminology - whether we have Reactor Trip on Pressurizer Pressure - Low which is available at the maximum J power output at which this particular event is evaluated, or Pressurizer Pressure - Low (Safety Injection) and provide the associated response time to ,

validate proposed T.S. values. j Response: The NA enter for the required response time of reactor trip upon SI actuation is consistent with the Bases which states that trip j functions not utiized in the FSAR transient analyses will have the ] { requirement indicate not applicable in Table 3.3-2 (Reactor Trip , i System Instrumentation Response Times). However, as stated in Table I

J J l 3.3-5 (Engineered Safety Features Response Times). The terminology j in Note e, Table 7.2.1-4, should be Pressurizer Pressure-Low ] (Sa f ety Inj ection) . This wording will be corrected in a future update of the FSAR. {

                                                                                                                               )
                                                                                                                      ,        I (Question Sa)

TABLE 3.3-3, Item 7g <

Applicable modes: The current T.S. proposes Modes 1 and 2#. Condition 2# is )

an invalid MODE since # identifies the P-1) interlock which can be manually ! effected only at approx. 1900 psig and which can on'ly occur in MODE 3, i.e. , d the condition should be 3#. The licensee should explain and propose. :I1 Please advise why this limitation at. MODE 2 [or 3)# is proposed and how it may

j relate to plant operating procedures in MODES 3 and 4 and whether this block is in conformance with regu:rcory requirements. S i i i Response: The defeat of auxiliary feedwater pump auto-start is accomplished by depressing a switch that is interlocked with the P-11 permissive. l ' Thus the auto-start can only be defeated below a pressurizer pressure of 1955 psig. However, the same defeat will prevent auto-start ou low-low steam generator level (Table 3.3-3, Item 7e(1)). Since this auto-start capability is required in MODES 1, 2, ] ' and 3, the defest switch is not used in these modes. Therefore the entry for APPE CABLE MODES on Item 7g is not important as it is l t controlled by the more limiting Item 7c(1). , The statement that P-11 can only occur in MODE 3 is not accurate. MODE 2 is defined as operation with T-avg. >350*F, K 0 power $5% RTP. Therefore, suberitical operation witIN >vg.99, a and 1350*F. l 1s in Mode 2 if K ff is not less than 0.99. Critical _ operation is i restricted to T-a,g. v 1551*F, but even then the pressure-temperature operating limits permit pressures below 1955 psig. As a practical { matter, pressure is maintained in the normal operating range (s2235 psig) during SODE 2. The 2# referred to in the question is retained l i to require that MODE 2 operation above F-11 is with the Item 7g , auto-start enabled, a (Question Sb) TABLE 3.3-3, Item 8 This is limited in Applicability to MODES 1, 2, 3 by the proposed T.S. Since a LOCA in MODE 4 is part of the Licensing Basis, see later section 3/4.5, ECCS under GENERAL, the licensee should evaluate the reasons for, and the consequences of, not proposing this OPERABLE IN HODE 4, and not being i available in MODE 5, to counter the consequences of potential LOCAs and loss of RHR cooling in these MOSES. The proposed T S. is non-conservative with respect to the Licesing Basis; the Licensee shall evaluate and propose. I s.

p ., l { 1 j l l Eesponse: This specification is consistent with other standard technical specifications which require operator action to mitigate the l consequences of a LOCA in these modes. , : (Question 6a) TAELE 3.3-4 Ite: 4d I The trip set point is currently specified at -100 psi /sec. Westinghouse Set i Feint Methodology for Unit 1, reference 18, shows this value to be "-110 psi"; an additional descriptor is also necessary reading: "with a time constant of l 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 conservatism, it is clear to delete the

                       - ir. -100 as the description reads: " Negative Steam Line Pressure Rate -

High so that T.S. values should read as 100 psi and 110 psi. This is also 3 ( internally consistent with the descriptor in Table 2.2-1, Item 4, namely: 4 Power Range, Neutron Flux High Negative Rate, 5% of RTP with a time constant ' of 2 seconds. Response: Since ne safety analysis limit exists for the negative steam line i pressure rate setpoint (i.e., it is not assumed in transient ' i analyses), the Setpoint Methodology.(Reference 18) listed the T.S. 1 values. The T.S. limits were revised at a later date and thus a ! discrepancy between the Reference 18 and T.S. values exists. f In order to correct a typographical error and adequately define the ! setpoint, a T.S. revision will be pursued in the following form: l Trip 03tpoint Allowable Value 4d. Negative Steam Line $100 psi 1120 psi ) Pressure Rate-High with a rate / lag with a rate / lag function time function time constant >50 constant >50. seconds seconds (Question 6b) TABII 3.3-4, Items 7c(1) and (2) 1 This technical specification provides that the motor-driven ATW Pumps start on l low-low in one SG whereas the turbine driven pumps require low-low in two l SGs. This appears to be in conflict with the accident evaluation in the Licensing Basis FSAR as elaborated below. [This however is not conflict with the Instrumentation & Control Logic of the FSAR.]

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

L _ _ _ _--_-- - _ - - . - - - - - - - - - - - - - - _ - . - - - - - - - - - - - --

d

                                                                                       >                                   v "The auxiliary feedwater system is actuated by the low-low Steam Generator                                                      i Water Level Signal. The auxiliary feedwater system is assumed to' supply                                                          i a total of 450 gpm to three intact steam generators.                                                                 ,

I

l Reference 5, Section 10.4.7.2.2 states that " Travel stops are set on the l

             -steam generator flow control valves.such that the turbine driven pump can                                                          l supply 450 gym to three intact steam generators while feeding one faulted generator and both motor driven pumps together can supply 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 steac generators".                                                                                                         l i
         *                                                                                                                                      \

Reference 7 Related Section 15.4.2.2.2, page 15.4-13a (revision 38), states: "The single active failure assumed in the analysis is the turbine driven auxiliary feedwater pump. The motor driven pump that is headered to the steam generator with the ruptured main feedline supplies 110 gpm tc the intact steam generator. The motor driven pump that'is headered tco j two intact steam generators supplies 170 gpm to each. This yields a i total flow of 450 to the intact steam generators one minute after reactor i 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 15.4-1 shows that after the accident is initiated at a programmed value of SG 1evel, the low-low SG level in the ruptured SG is reached at 20 secs. later, and ! aaxiliary 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 turbine driven pumps are initiated (with the single failure being in the turbine driven pumps). This is not in accord with the T.S. If it is assumed that low-low level in the other SGs is also reached at the same time by bubble collapse, please justify. We note that the Reactor

       & Turbine Control System is designed so that under normal operation, collapse of SG level on Turbine 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.

The proposed T.S.s Item 7c(1) and 7c(2) appear to be non-conservative in respect of accident analysis used'in the Licensing Bases. The licensee shall clarify, evaluate and propose. l'

Response

It appears that the question is "Since one motor-driven pump supplies 110 gpm to an intact generator and the other otor driven pump supplies 170gpm to intact generators, where does the remaining " 170 gpm (450 - 110 - 170), supplied to the intact generators, come from if not from the turbine-driven pump?". The new FSAR Chapter 15 analyses for optimized fuel make clear that the "two motor-driven pumps together deliver 450 gpm to the three intact steam generators allowing for spillage out of the break (Section 15.2.8.2, page 15.2.8, 1984 Update). To clarify exactly the analysis assumption - One meter driven auxiliary feedwater pump

   '-                supplies 110 gpm to an intact steam generator (the remainder spills out the break in the f aulted loop) and the other motor driven pump supplies 170 gpm to each of the other two intact steam generators, this totals to 450 gpm.                                                      !

If the failure of a motor driven pump is assumed, the turbine ) i driven pump alone would supply at least 450 gpm to the intact loops. The turbine driven pump is actuated on low-low-level in at/least two steam generators. It is assumed that low-low level is : reached in the other (non faulted) steam generators as a result of the bubble collapse following turbine trip when the low-low level ) i reactor trip is actuated from the faulted loop. This occurs because i for this accident condition (i.e. not normal operation) the mass ; icventory in the intact steam generators is reduced significantly [ prior to reactor trip on low-low level in the faulted loop. The j sbrinkage caused by bubble collapse from this reduced mass q condition would cause low-low level to be reached in the other steam generators. i i l 4 (Questler. 6c) TABLE 3. 3-4 Item 9 1 Confirm the bases for the set points and allowable values specified. Response: The bases for the setpoints and allowable values-specified are to I ensure Auxiliary Feedwater capability upon loss of power while l minimizing the possible initiation of the sequence with the voltage greater than the limits of associated motors. 1 l. I 1 (Question 7a) l TABLE 3.3-5, Item 2a A value of 5 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 upon for diverse protection. However, it is the only automatic initiation of Safety Injection protection below [P-ll). Other events dependent upon a SI generating signal, particularly circumstances described under Items 3a and 4a below, shows offsite secs (without safety power). analyses limits of 5 12 secs (with offiste power) and 1 22 At this time, the proposed T.S. value is less conservative than others used in Safety Analysis. accordingly. The licensee shall evaluate this difference and propose l l L

Response: The entry for Table 3.3-5, Item 3a is identical to Item 2a for the loss of off site power case, i.e., each is 27 seconds. As explained in the Notes for Table 3.3-5, the difference between Item 4a and items 2a and'3a is that 4a does not include a delay for the RER pumps to attain'their discharge pressure. 'This is appropriate since Item 4a deals with steam line break protection, as opposed to LOCA protection. The RHR pumps, although started for a steam line break, are not expected to deliver flow because of the higher RCS pressure. Therefore, the additional 5 second delay for these pumps to attain their, discharge pressure is not relevant to ESF response time for this actuating signal. l (Question 7b) TABLE 3.3-5, Item 2b The descriptor (From SI), should be deleted as it is incorrect. The response time given is < 2 secs and this different from the FSAR, Reference 5, page 7.3-6 which gives'a maximum time of I sec. This value is less conservative than the FSAR and the licensee shall evaluate and propose accordingly. Response: The descriptor "(From SI)" is correct in that the allowable delay for a reactor trip due to the SI actuation signal is 2 seconds. This value is independent of the setpoint and associated delay of the initiator of SI. The reference 5, page 7.3-8 maximum time of 1.0 second is the licit on the delay associated with SI actuation upon exceeding the high containment pressure setpoint. No credit is taken for reactor trip signal resulting from safety injection signal in any LOCA analysis. In the McGuire Unit 1 initial core large break LOCA analysis no credit is taken for reactor trip (rod insertion) at all. In the McGuire Unit 1 initial core small break LOCA a low pressurizer signal causes the reactor to trip. No credit for the control rods is taken until they are fully inserted. (Question 7c) TABLE 3.3-5, Item 2d The proposed T.S. values tre 1853) (with offsite power) and 28(4) without offsite power. Reference 5, page 7.3-8 shows that initiation of ESFAS from this source is I 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 secs seems feasible, plus 10 secs giving 26 secs total without offsite power. l l

The propcsed T.S. values should be checked against those used as Safety Analysis lim:ts for related Conditions II, III, and IV occurrences using SI. Values used by licensee shall be provided, compared with Item 2d, and any differences evaluated. Response: Following a design basis large LOCA, the isolation valve closure time depends upon the time when fuel failure occurs and fission products are released to the containment environment. The only isolation valves explicitly considered in the radiological consequences analysis of a LOCA are those in the containment purge and pressure relief lines which connect containment to the environment. For isolation valves in lines filled with process fluid a relatively long time is needed for the associated piping system to drain of fluid and expose the valve seat to the containment gases or for activity to migrate, due to the-concentration gradient, through the process fluid and out the isolation valve. Hence, as long as isolation valve closure times for process lines are short (less than'1 min. per ANS 56.2) they. need not be modeled in the dose calculations. (Question 7d) T ABLE 3. 3-5, I tet 2e This is given as N.A. This is not so; response times have been used to minimize offsite consequences of any Condition occurring whilst containment purge and exhaust is beicg used. This proposed T.S. is less conservative than the licensing basis. The license shall evaluate and propose. Response: Section 15.B.2 of the McGuire FSAR considers the caso of a LOCA concurrect with lower containment pressure relief. The results of the additional offsjte dose due to this accident are presented in table 15.0.11-1. One of the parameters used to evaluate this case is the isolation time for the Containment Air Release and Addition (VQ) System valves which are used in venting lower containment. 1,51e 15.B.2-1 indicates the isolation time for these valves is 4 secsnds. Section 9.5.12.3 indicates that these valves auto close on a containment isolation, and that they have a 3 second closure time A technical specification revision to show a response time of < 4 seco,ds for this item will be pursued. This would be consistent with the allowable I second for generating an ESF response as indi ated oc fige 7.3-8 of the McGuire FSAR and the 3 second valve clos.ng tiae es tedieeted above.

IOues :en 7ei I IAELE 3.3-5. Item 2f 1 The licensee proposes N.A. but earlier review shows ATV initiation on. Containment 'i Fressure-High and especially in MODES 3 and 4. This is less conservative than { the licensing basis; the licensee shall evaluate and propose. ; Response: No credit is taken for ATW flow being initiated from a Containment Pressure - High signal in analyses. j (Question 7f). Y TAELE 3.3-5, Item 3a Values of 1 27 )/12(3 secs are proposed. Reference 5, page 7.3-8, shows a maximum initiating time of ESTAS 1.0 secs from this signal. i The value of 12 secs (with offsite power) is consistent with safety analysis limits giver. for the MSLB in reference 7, page 15.4-10, Section 7 where "In 12 seconds, the valves are a sumed 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 within the licensing basis of 12 secs. The proposed value of 27 secs (with LOOP) is however larger than the value of 22 seconds f rom the reference described above (i.e. ,12 sees + 10 sets delay for start of diesel). This value of 27 secs therefore appears less ; conservative than the FSAR, reference 7, page 15.4-10, and the licensee shall evaluate and propose. ] j Response: This question is related to the question on Item 2a. For a steam line break the RHR pumps are not expected to deliver inventory and the additional 5 second delay for them to attain their discharge pressure is not included in the safety analysis. (Question 7g) TABLE 3.3-5, Item 3b The descriptor (from SI) is incorrect and should be deleted. A value of 1 2 secs is proposed. The FSAR in Reference 5, page 7.3-8, quotes a value of 3 1 secs. The proposed T.S. value appears less conservative than the Safety _ Analysis Limit and the licensee should evaluate and propose.

p

1 i e Response: The descriptor "(from SI)" is correct in that the allowable delay for a reactor trip due to the SI actuation signal is 2.0 seconds. This value is independent of the se point and associated delay of the initiator of SI. The Reference 5, page 7.3-8, maximum time of 1.0 second is the limit on the delay associated with SI actuation upon exceeding the Press rizer Pressure - Low setpoint. The chapter 15 safety analyses do not take credit for a reactor trip from an SI signal initiated by low-low pressurizer. (Ref. Question 7b Response). (Question 7b) i TAELE 3. 3-5, Itee 3d i The proposed T.S. is 5 18(3}/28(4} secs. Reference our comments and requirements under Item 2d above. kesponse: Reference our response under item 2d above. l (

(Question 71) j TABLE 3. 3-5, Ite: 3e l { The proposed T.S. is NA. Reference our comments and requirements under 2e. above.

Response: Reference our response under Item 2e above. (Question 7j) TABLE 3.3-5, Item 3f The licensee proposes NA (not applicable). Safety injection logic closes the main feedwater isolation valves for every event 3.3-4, Table in which SI is initiated proposed Item c). (reference earlier sections of this review Therefore, every such event initiated by a SI initiator must be analyzed with a restoration of ATW and a related response time. It is outside the licensing basis to not propose a value for this response time. This T.S. value is therefore non-conservative; the licensee shall evaluate and propose. Response: The only non-LOCA transient which assumes ESF actuation on Pressurizer Pressure Low-Low is the Main Steamline Depressurization event (Inadvertent Opening of a Steam Generator Safety, Relief, or Dump Valve). For this event it is conservatively assumed that

I l auxiliary feedwater is actuated at the maximum flow rate at the-initiation of the event to accentuate the cooldown. Any delay in auxiliary feedvater actuation would be beneficial and therefore a respense time requirement is not applicable or appropriate. (Question 7k) TABLE 3.3-5 Item 4e s The proposed T.S. is NA. Reference our comments and requirements under Item 2d above, Response: I Reference our response un' der Item 2e above. (Question 71) ! TABLE 3.3-5, Item 4h The proposed T.S. value is 1 9 secs. Reference 5, page 7.3-8 states that the maximum allowable times for generating 1 steam break protection are (1) from steam line pressure rate, 2 secs, and (:) from steam line pressure-low, 2 secs. Further, Reference 7, page 15.4-6 i states that the fast acting steam line stop valves are " designed so close in 5 secs. .". A minimum closure of 7 secs seems likely. ; For actual safety analysis limits, Reference 7, Table 15.4-1 (1 of 4) and 15.4-1 (2 of 4) both show a difference of seven (7) secs between arriving at the " Low Steam Line Pressure Setpoint" and "All Main Steam Isolation Valves Closed." [In the case of Feedwater System Pipe Rupture]. The proposed T.S. value of 5 9 secs is therefore greater than the Safety Analysis Limit.

The proposed T.S. must therefore be considered less conservative for this event. The licensee shall evaluate and propose.

Response

Item 4h in Technical Specification Table 3.3-5 has been changed to l a limit of i 7 seconds (Ref. Amendment nos. 29 (Unit 1) and 10 (Unit 2)). (Question 7m) TABLE 3.3-5, Item Sa [31; Licensee shall provide the Safety Analysis Limit and compare with the proposed value of 1 45 secs. Evaluate and propose as necessary, i

                                                                              } '.

4: Festcase: The response time for containment spray following a high containment pressure signal is specified'at 45 seconds in the McGuire Technical Specifications. This value is consistent with the FSAR containment analysis actuation assumption as shown in FSAR Table 6.2.1-13c. Event times from the McGuire limiting case break

 ,                                               mass / energy release analysis are reported in Table 6.2.1-29; the time of spray actuation has no effect on the mass / energy releases calculated.

(Question 7n) TABLE 3.3-5, Itec 6b The proposed T.S. is 5 13 sees. 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 1 5 sees for Main Feedwa te r ' Containment Isolation and 310 secs for Main Feedwater to Auxiliary Feedwater Isolatioc. i A total time to isolation of MTV of 5 13 secs seems appropriate to available equipment. i However the current safety analysis depending on this response time is that for the Excessive Cooldown occurrence under Reference 7, page 15.2-28, and for this, no value is quoted for icolation of main feedwater which'is the initiator of the event. However, Figure 15.2.10-2 shows that with initiation 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 is' expected to take another 13 secs to actually isolate, we could assume an additional mass increase of another 13% to give a total of approximately 1.48 the initial value. The above additional Main Feedwater level can affect the consequences of the event at power, if there has been a trip, with a potential for power i restoration and/or overfill of the SG to cause water ingress into the main l steam lines. Additionally, it can have consequences of potentially larger importance for the event occurring from suberitical zero power. l Reference also our concerns under item Table 3.3-4, Items 11b and 11a above. The licensee shall evaluate the related concerns, including the extended MFW valve isolation times, to determine their safety significance, and propose as required. Until that time, it must be concluded that since a zero (0) value has been used in the current analysis, the licensee has a potentially non-conservative situation with respect to regulatory requirements of reactivity control and regulatory concerns for flooding of the main steam lines.

<m l I i I l t hesponse: Excessive Feedwater Flow at Full Power is analyzed in Section 15.1.2 of the ."cGuire FSAR. Table 15.1.2-1, page 1 of 2, 1984 Update, giv.es the sequence of events for this analysis. The High-High SG Level setpoint is reached at 27 seconds with feedwater isolation occurring 9 seconds later. This 9 second value agrees with the values used for feedwater isolation on Safety Injection. To be consistent with the current safety analysis the Technical Specifications value for item 6b of Table 3.3-5 should be 19 seconds. Another alternative is to reanalyze the Excessive Feedwater Flow event with the longer delay time. Duke will pursue a Technical Specification revision or reanalysis. 1 (Question 7o) TABLE 3.3-5, Item y Response time proposed as 1 60 secs. The licensee shall provide the bases for this value, evaluate against this 5 60 secs, and propose as necessary. Response: The automatic switchover to recirculation is initiated when the level setpoint in the RWST is reached. The setpoint determination includes allowances for switchover delay > 60 seconds and plant procedures test to ensure switchover delay 5 60 seconds per Table 3.3-5, Item 12. l General Response to Questions Sa-Se: These questions in general deal with the conservatism of the FSAR Chapter 15 safety analyses for events initiated from HODES 3-5. Specifically the question of the number of RCS loops in operation, for heat removal or other purposes, appears many times. Since the McGuire Technical Specifications and Westinghouse Standard Technical Specifications are identifical for MODES 3-5 for T.S. 3.4.1, Reactor Coolant Loops and Coolant Circulation, any questions regarding these matters should be resolved on a generic basis and are not specific to McGuire. ! Therefore, with items which are specific the responses to each question will, deal only to McGuire. (Question Ba) SECTION 3/4.4.1, G.2.6.1 OCCURRENCES WITH RAPID REACTIVITY INCREASE Concerning " Uncontrolled Rod Cluster Control Assembly Bank Withdrawal from Sub-critical Condition." Current docketed analysis in reference 7, Section 15.2.1, page 15.2-2 is based on four operating loops. This event is possible down to and including Mode 5. Current FSAR analysis trips the reactor on Power Range, Neutron Flux Low

l Setpoint (25%) at a Safety Analysis Limit of 35% (reference page 15.2-3, Item 3). Tce principal determinant of ultimate power level is Doppler coefficient; contribution of moderater reactivity coefficient is negligible (reference

page 15.2-3, Items 1 & 2). The event is initiated from hot zero power j.

(reference 7, page 15.2-4, Item 3). 4 RCS pumps are operating. ' Given the circumstances cf the proposed T.S., any T.S. allowing OPERABILITY of j less than 4 RCS Loop in MODE 3 would be in nonconformance with the current FSAR i in a nonconservative manner, and the licensee would be required to evaluate and propose. Furthermore, increased boron concentrations would not change this requiremen . Additional events of a similar nature, with a rapid increase in reactivity include: a) Uncontrolled Boron Dilution (reference 7, page 15.2-13). 1 b) Startup cf an Inactive Reactor Coolant Loop (reference 7,'page 15.2-19, revision 7). j c) Excessive Heat Removal Due to Feedwater System Malfunction (reference 7., page 15.2-30, revision 7) concerning ir.itiation with'the reactor at zero power). Until the licensee clarifies availability of MFW during MODES 3-j through 5, this must be considered a potential. occurrence. I d) Single rod cluster control assembly withdrawal (reference 7, Page 15.3-9, revision 7). Although the Licensing Basis is at 100% power, the ; circumstances from zero power should be reviewed. e) Rupture of a Control Rod Drive Mechen sm Housing, at Zero Power (reference 7, Page 15.4-30; revision 42). 1 f) Major Rupture of a Main Steam Line (see below). i Response: No McGuire specific concerns are ra3 sed in this question. Refer to the general response to Questions 8a-8e. . l I (Question 8b) ! SECTION 3/4.4.1, G.2.6.2 STEAM LINE' BREAKS Concerning " Major Rupture of a Main Steamline." l This Event is discussed in Accident Analyses in Reference 7, Section 15.4.2 and Reference 8, Item 212.75, page Q 212-47d & e, Item 25. Reference 8 !' proposes that the resulting impact on shutdown margins from this event during MODES 3, 4, and 5 are improved over that of the design basis (hot zero power, just critical, Tavg = 557') as:

                                    " Operating Instructions require that the boron concentration be increased to at least the cold shutdown boron concentration before cooldown is initiated. This requirement insures a minimum of 1%

Ak/k shutdown margin t -

r

 ,               at a Reactor Coolant System temperature of 200'F. This condition assures that the minimum shutdown mergin experienced during the                          ,

streamline rupture from zero power shown in the safety analysis is less than-the case where safety injection actuation is manually blocked on low steamline pressure and low pressurizer pressure." ! This position gives no measure of the resulting shutdown margins and/or power level and, the consequences of a stuck rod, with only 2 RC loops operating instead of four. It is-conceivable 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 ic'the core due to circulation , (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, a MSLB cools the RCS down to 212'F e.g., the residual shutdown margin will be l*s 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 at what-condition during cooldown the 1.6*A delta k/k is reached. 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-conformance. with the current Licensing Easis Safety Analysis in the FSAR in a ocn-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 en Power Flux Trip, but on Pressurizer Pressure-Low (SI) (above P-11) [ reference our r-quired confirmation of this in an earlier item] so the Power Flux Trip is not required to be Operable, j At less than P-11, these circumstances are changed for the MSLB, and reactor trip does not occur until Containment - Hi is achieved, for a break inside containment. For a break outside containment, however, high negative steam rate isolates main ' steam isolation valves only, but there is no Safety Injection, no Reactor Trip ! (on SI), and under the existing proposed !.S. no safety related Reactor l Trip System Instrumentation of any nature to trip the reactor and insert the 1 movable control rods to benefit from potentially increased available shutdown margin. In addition to all this, the licensee proposes that MSIV closure times under these conditions is Not Applicable. Given the circumstances of the proposed T.S., the T.S. allowing OPERABILITY of less than 4 RCS Loop in MODE 3 under these circumstances would be in nonconformance with the current Licensing Basis FSAR in a nonconservative manner, and the licensee would be required to evaluate and propose. Additional events which exhibit a rapid cooldown and depressurization of the l RCS; are: a) Accidental Depressurization of the main steam system at no load, (reference 7, page 15.2-35, revision 36). ,

l ). l l i Minor Secondary System Pipe Breaks (at no load]; reference 7, page 15.3-4, revision 27). Response: Changes in the Technical Specifications and plant procedures have ) occured since the DP0 questions were posed (boration to cold shutdown prior to starting cooldown is no longer required). The required shutdown margin for RCS temperature above 200*F is 1.3% ak/h. The shutdown margin requirement for temperatures equal to or less than 200'F is 1.0% Ak/k. Variations in initial condit.ons for the steamline break transient were analyzed in WCAP-9226 and support the conservative assumptions in the FSAR analysis, i Closure times for the Main Steam Isolation Valves (MSIVs) are implied in the Technical Specifications. In Table 3.3-5, Items 4h, Sc, and 8, response times are given for the Steam Line Isolation function. This time includes the MSIV closure time. Other concerns raised in this question are generic. ' Refer to the general t response to Questions 8a-8e. (Question Sc) SECTION 3/4.4.1, G.2.6.3 LOSS OF PRIMARY COOLANT Concerning: "Small Break LOCA". if This is discussed in reference 7, Section 15.3.1, for a SBLOCA from rated power, and reference 8, Item 212.75, page Q 212-47b for a SBLOCA between RCS conditions of 1900 psig and 1000 psig/425'F in Hot Standby, and Q212-64, Item 3 together 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 Standby to and including RHR. l 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 pumps through the core as they are coasting down following trip:   therefore upward flow through the core is maintained."                          l Topical Report, 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 a) all pumps 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 from the flow reversal compared to the case } } where pumps are immediately tripped." and "The net result of these effects is a

1 1 staller peak clad temperature for the pumps running case compared to the pumps tripped case. Hence, for ECCS analyses for W 4 loop plants the reactc-r coolant pumps are assumed to be tripped at the initiation of a postulated LOCA and a locked rotor pump resistance is used for reflood." At this time therefore, the b'RC must conclude that RCS pump operation and ccastdown is important in reducing the loss of core level subsequent to the event; also in maintaining unseparated two phase flow conditions and in ensuing rapid boron (mixing and) injection to the core. Rapid boron injection would not be an important issue if boron concentrations are already at cold shutdown values, but minimizing 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 pumps OPERABLE in MODE 3 down ,to 425 psig/350*F. The current proposed T.S. are therefore nonconservative and the licensee must evaluate and propose. Given the circumstances of the proposed T.S., operability of less than 4 RCS-loops in MODE 3 would be in non-conformance with the current Safety Analyse's Limits in a nou-conservative manner and the licensee is required to evaluate and propese. Additional events of a similar nature to the S3LOCA events include: a) Accidental Depressurization of the Reactor Coolant System (reference 7, 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, 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, the Power Range Eeutron Flux, High Setpoint trips the reactor; Pressurizer Pressure-Low (SI) initi~ates Safety Injection; reference 7, page 15.4-29, revision 43, paras. I and'5. Whereas both these projections are proposed by the T.S. in MODE 2, they are not proposed for MODE 3 which differs from the circumstances of NODE 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 No of RCPs than 4; it has been analyzed for 2 only. This is an accident with substantial 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 nonconservative with respect to the Licensing Bases. The licensee shall evaluate and propose.

I

                                                                                                                                                                            'I l

C:ncerr.ing 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;. ir. Reference E, Item 212.75, page Q 212.47, for a LOCA between RCS conditions I cf 1900 psig and 1000 psig/425'T in Hot Standby; in Item 212.90 (6.3), page 212-61, for a LOCA at and less than 1000 psig/425' in Hot Standby, and on page Q 212-61b, Item 29 for a LOCA in the RHR Mode at 425 psig/350*F. l f l as for the small break LOCA, these analyses are presumably based on 4 RCS loop J operation, with in general, loss of power to RCS pump' on Safety Injection. j 1 The large break LOCA analyses used the Topical Report WCAP-8479, reference 7, I page 15.4-1. At this time, we expect no difference in the importance of RCPs tc 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 and small break LOCAs. (

-)

J Given the circumstances of the proposed T.S. , any T.S. allowing OPERABILITY of fewer than 4 RCS loops in MODE 3 would le is nonconformance with the Licensing , Bases FSAR in a conconservative manner, and the licensee is required to evaluate and propose. Fesponse: No McGuire specific concerns are raised in this question. Refer to the general response to Questions 8a-8e. (Question 8d) SECTION 3/4.4.1, G.2.6.4 OCCURRENCES CAUSING AN INITI AL INCREASE IN RCS TEMPERATURE Those events causing increases in RCS temperature are of concern because of j the potential influence of the positive moderator temperature coefficient resulting from the increased boron concentration. These could be: a) Main Rupture of a Main Feed Line (Reference 7, page 15.4-10, revision 30), although this is normally evaluated at Rated power with no provision for < evaluation at zero power. b) Startup of an Inactive Reactor Coolant Loop. ) h c) Loss of Offsite Power (reference 7, page 15.2-19, revision 7). d) Partial Loss of Forced Reactor Coolant Flow (Reference 7, page 15.2-16, revision 7). e) Complete Loss of Forced Reactor Coolant Flow (Reference 7,'page 15.3-7, revision 7). - Except for item b; all these events are licensing bases events from rated j power, and not zero power, so that their importance would normally be 1 minimal except for th,e positive Moderator Temperature Coefficient and the 1 l complete lack of safety-related Reactor Trip protection proposed with the j Reactor Trip System Instrumentation T.S. At this time we see no protection 3 against positive temperature coefficients in MODE 3 [4, 5, & 6]. ' l

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I 1

                                                                                      ,                  i Given the circumstances of the proposed T.S.. operability of less than 4 RCS locps in MODE 3 would be in nonconformance with the current Safety Analyses                       j Limits in a nonconservative manner. The licensee is required to evaluate and f

propose. l Response: No McGuire specific concerns are raised in this question. Refer to the general response to Questions 8a-8e. ' ; q (Question Sc) T.S. 3.

4.1 CONCLUSION

S Occurrsace II. III and IV Events in MODES 3, 4, and 5 can result in returns to power with high peaking coefficients requiring effective reactivity control j and/or reacter core flow for RCS protection, including DNBR, at the very j substantially reduced pressure levels in the loop [2250 psf.g to 425 psig and lesr}. Concomitant decreases in RCS temperatures are beneficial, but t ] importance of RCS pressure may be dominant. Acceptable RCS protection therefore requires RCS flows which are substantial, and/or effective reactivity control including combined action to limit potential reactivity excursions. At this time, with the proposed T.S., 4 RCS loops (with increased Reactor Trip Protection) would be required at entry into and during MODE 3 to meet the requirements of just the Licensing Basis Events From Zero Power. In MODE 4, operation of 4 RCS Loops, whilst on RHR, may be undesirable because of the substantial additional burden on the RER 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. An additional available alternate action is to use, within MODE 4, a l 1 minimum set of RCPs (and~ loops) as established by Safety Analysis, to cool the plant down to effectively zero pressure (gauge) in the Steam Generators (or less if the condenser was still available] before transferring the heat sink to the RER system. This would ensure control of steamline break, and LOCA events, small and large, down to conditions where RCS flows are not nece s sa ry. The current T.S. are nonconservative in respect to the Licensing Basis in respect to these concerns. The Licensee shall evaluate and propose. Response: No McGuire specific concerns are raised in this quastion. Refer to the general response to Questions 8a-8e.

~,

1 (Question 9)

      !.5. Page 3/    4-2 Earlier concerns under General 2.6.1 addressed the need to evaluate the consequences of the startup of an inactive Reactor Coolant Loop in this MODE.

Ne apparent T.S. provision has been provided in the proposed T.S. The 12censee shall evaluate and propose. ACTION b. states: ' 1 "With ne reactor coolant loop in operation, suspend all operations involving a reduction in boton concentration of the Reactor Coolant ! System and immediately initiate corrective action to return the required I reactor coolant leop to operation." This instruction is invalid. >The only Licensing Basis actioc.available is the Emergency operating Guidelines for natural circulation. This proposal is ) nonconservative with respect tc the Licensing Basis. The licensee shall ' evaluate and propose. Response: The actions included in. ACTION b. are '1) suspend deboration operations and 2) immediately initiate action to restore forced circulation. The actions are obviously, valid responses to the condition. There is no Emergency Operating Procedure at McGuire for natural circulation. - There is Abnormal Procedure AP/1&2/A/5500/09, Plant Operations During Natural Circulation, which addresses the initiation, verification, and maintenance of natural circulation. This procedure would be implemented under this condition. l

(Question 10) 7.S. Page 3/4 4-3 The licensee shall evaluate as outlined earlier under item, General, for RCS loops operability requirements and make proposals relative to the status of many l elements of the protection and operations system to ensure that RCS safety is I maintained for related Condition II, III and IV occurrences. At this time, with the proposed T.S. in which limited boration is used and Reactor Trip System , safety related instrumentation and Safety Injection instrumentation are all ' j but eliminated, the safety status of the facility is outside the Licensing l Basis of the FSAR in a nonconservative manner. Each of the OPERABLI loops, whether RCS _ or RHR, are to be energized from separ-ate power divisions to protect against single failure of a bus or distribution i system. When the RCS systems are used, the related Auxiliary Feedwater Systems are also required to be operable.

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The addinonal requirement proposed, for two RCS loops to be operable whenever REE loop /s are in operation, is based upon reference 8, page Q 212-55 and 56,. i to provide for the failure of a single motorized valve in the RHR/RCS suction l line in b th MODES 4 and 5 and the possible non-availability of offsite power I sources. The FSAR provides, that on failure of the valve: { l "Approximately 3 hours are available to the operator to establin,h an ! alternate means of core cooling. This is the time it would take to heat I the available RCS volume from 350 F to the sturation temperature for 400 psi (445 F), assuming the maximum 24 hours decay heat load. To restore core cooling, the operator only has to return to heat removal via the steam generators. The operator can employ either steam dump to the main condenser or to the atmosphere, with makeup to the steam generators from the Auxiliary Feedwater System. The time required to establish the alternate means of heat removal is only the few minutes necessry to open the steam dump valves and to start up the Auxiliary Feedwater System." The applicability MODE 4, is necessarily qualified by (less than 425 psig/350 F} by the LOCA analyses already referenced above under our Review l Section 3/4 4.1 Subsection G.2.6.3 "concerning Large Break loss of coolant accident." See Reference 8, page Q 212-47d where it is described'that "Af ter several hours into the cooldown procedure (a minimum time is ! approximately four hours) when the Res pressure and temperature have ' decreased to 400 psig and 350 F." And arising from a later revision 25, the FSAR Advises on page Q 212-61b Revision 29 concerning ECCS calculations in a later submittal under Revision 26 that "The response provided in Revision 26 addressed the subject of operator actions and ECCS availability. Consistent with the information provided 1 in Revision 28, a postulated LOCA in the RHR mode at 425 psig RCS i pressure has been assessed." l Surveillance requirement 4.4.1.3.2 should verify SG water level at the Safety i 1 Analysis Limit fo.r the Licensing Basis, which is the no-load programmed level, not the current proposed T.S. valve which is the S.G. Low-Low Level [ Reactor Trip) and ATV actuation. This proposed T.S. is nonconservative with respect to i the current Safety Analysis Limits and the licensee shall evaluate and propose. Surveillance requirement 4.4.1.3.3, verifying one loop in operation every 12 hours, is insupportable 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 RRR has been provided by the FSAR under reference 8, page Q 212-56, Item:

           " 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)."

 *b     5:r.:e currently, these two types of alarms are the only means of alerting the y erator to a loss of flow condition in the loop, which is beyond the Safety Analysis limits, the alarms on both the RCS and loop flows should be safety-related and included within the T.S.; and without further analysis at this time, twc loops should.be placed in operation. A proposal is made by the NRC for low          l flow alarms in each of the separated cooling systems, under proposed T.S. page 3/4 4-6a of this rev!2w. Regular surveillance should be proposed to ensure             ) l that they remain operable as appropriate, over a specified surveillance                  '

period. The Surveillance requirement, every 12 hours is intended to ensure not only I that the system is operating, but that it is operating at process conditions which can be evaluated to show that the equipment is capable of performing its ! design basis Safety Function. The current surveillance requirements for this itec, i.e., for the RCS and RER systems in Hot Shutdown in T.S. Item 4.4.1.3.3, are absen't this information; it is therefore nonconservative and the licensee shall evaluate and propose. Item 4.4.1.4.4 (Proposed). It is proposed that an additional item be inserted which reads: "The related auxiliary Feedwater System shall be determined OPERABLE as per the requirements of T.S. 3.7.1.2 [and 3.7.1.2.a as applicable) ." ' Current proposed T.S.s on T.S. page 3/4 7-4 are nonconservative in this matter by not providing any operability requirements for AFW in this MODE. The licensee shall evaluate and propose. I An additional item is also required in which Atmospheric Dump Pe.ves operability ' ' is established. The current T.S. are nonconservative 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 SG water level, AFW and atmospheric dump 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). Response: Several separate questions are raised here. The McGuire specific ones are answered as follows:

1) Each RHR train is powered from a separate.4160V bus in the Essential I Auxiliary Power System. Each reactor coolant pump is powered from a j separate 6900V bus in the Normal Auxiliary Power System.. !

l

2) It should be noted that the requirement of maintaining a specific level in the steam generator to verify operability was imposed by ;

the NRC and has no firm basis within Westinghouse. However, for an RCS loop to be operable, sufficient inventory is required in the secondary side for heat removal. In MODE 4 this can be assured by keeping the tube bundle covered. A reasonable way of ensuring this .; is to require that the secondary side level indicates within the l narrow range span. Accounting for errors, an indicated level at the low-low level setpoint assures that the level is at least at the bottom of the narrow range span. i l l

1 3 The safety analysis limit for reactor trip on lo-lo SG level is a i functicn with a value of 0% at no-load conditions. Adding allowances for reference leg heatup and instrument ertor gives the ! value of 12% used as the T.S. trip setpoint. The T.S. value is therefore conservative with respect to the safety analysis limit. l 3) The lov flow alarms oc the RER loops are to alert the operator to insufficient flow under RER conditions. They have no relation to the low flow reactor trip which inserts the control rods to control reactivity during low flow conditions at power. Boron is employed for reactivity contrcl in_the shutdown modes while rod insertion is impossible (if the rods are already inserted) or unnecessary (because of the boration). The current surveillance 4.4.1.3.3 requires verifying one RCS or j RHR loop in operation at least every 12 hours. The concern raised i apparently centers around the assertion that core cooling could-be ! lost without the knowledge of the operator since no protective functions or alarms are required to be operable by the technical f { ( specifications. However, it is expected that there would be { multiple indications of any problems that could cause a loss of j coclant loop. Although the appropriate alarms are not required by I the technical specifications to be operable, there is no reason to believe that all relevant alarms and other indicators would be inoperative during this mode. The otber issues raised in this question are not specific to McGuire. Refer to the general response to Questions 8a-Se. (Question lla) T.S. SECTION 3/4.5 At less than 400 psig and 350*F, the operator aligns tre Residual Heat Removal System. The valves in the line from the RWST are closed. 1

Response

This " question" is merely a statement of operator action to align RHR. It remains true and requires no response. (Question lib) T.S. 3.5 Below 400 psig, the system is in the RER 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.

r Fesponse: This " question" is merely a statement of operator action to align the s ECCS for use from a shutduwa coadition. It remains true and requires _{ no response. 1 i (Question 11c) T.S. 3.5 l l 4 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 has l been assessed. The initial conditions would be reached four hours after reactor shutdown. The integrity of the core after a postulated LOCA is l 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 i under an assumption of ten minute delay for operator reaction time. Assumptions: i { (a) The system pressure essentially reaches equilibrium with containment by { the time the volume of water above the bottom of the hot legs is 1 removed. (b) Upper plenum fluid volume between the top of the core and bottom of hot . legs is the only upper plenum fluid considered. (c) Volume between the core barrel and baffle is conservatively neglected. (d) 120% of the ANS decay heat curve for four hours after shutdown is utilized. l ) l Using the void fractions developed from the Yeh correlations and utilizing a hydrostatic pressure balance, the height of the steam-water mixture in the upper j plenum was generated. Incorporating the plant geometry, the total liquid mass 1 in the downcomer, core, and upper plenitm was calculated, i.e. . .a mass-initial I condition. Again by hydrostatic pressure balance, the height of ligttid in the I downcomer when the top of the core is just about to uncover was calculated. This infonnation along with core volume is used to develop a mass-final condition. That is, the mass is liquid 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 compared to the ten minute assumption for operator reaction time.

                    " Utilizing the preceding approach, the time calculated to just initiate an uncovery of the core is 13 minutes. The conclusion ir that even for the                                                                           3 conservative method outlined above, there exists adequate margin to retain a safe core condition even in relation to a ten minute operato.r-response-time assumption."                                                                                                                                     ;

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   - _ . _   a __a_  _ -__m  ____ - _ _ _ . _ _ _ - - _ _ _ _ _ _ _ _ _      m_ _ _ _ - _ _ _ _ _ _ _ _ _ -_ . - .m   - _ _ - , _ _ _ _ _ _ _ _ . _ _ ~ - _

n. . .. . . .. .. . . .

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               -These operater requirements are verified, in general,- by reference 12, SER Supplement 2, page 6.6-6.8, under " Emergency Core Cooling System - Performance.          I Evaluation", and pages 7-1 and 7-2 under " Upper Head Injection Irelation Valves".

Additionally, the status of the ECCS systems from entry into the RRR MODE j through cooldown, i.e., from 425 psig/350*F through MODE 5 is clarified by the l fcllowing extract from reference 11, suppl. SER No. 1, pages 5-1 and 5-2 which ' confirms continuance of the alignment at the end of MODE 3 425 psig/350'T ! through both MODES 4 and 5. Response: This " question" is largely a quotation from the FSAR. The last

two paragraphs, while not from the FSAR, are simply statements

introducing a quotation from the SER. Therefore, this requires no response. 1 (Question 12a) T.S. 3.5.1.1.d. 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 and drift, should not this value be a higher setpoint of approximately 450 psig? The specified setpoint values proposed in the T.S. of 400 to 454 psig can therefore give actual values which are lower than in the Licensing basis FSAR and be non-conservative. The Licensee shall evaluate and propose. Response: The bases for the T.S. 3.5.1 limit of Cold Leg Accumulator cover pressure of between 400-454 psig is the assumed value in the LOCA analysis (FSAR Chapter 15). Allowance for channel error and drift are accounted for in the determination of the T.S. requirements. The numbers in Table 6.3.2-1 are nominal and minimum values as required by T.S. 3.5.1 and are in agreement with the T.S. 3.5.1 limits. Recent Technical Specification changes (Ref. unit 1/2 License Amendments 57/38) associated with the removal / isolation of the UHI System involve revising the : Cold Leg Accumulator cover pressure to between 585 and 639 psig.

j (Question 12b) T.S. 4.5.1.1.1.d.1 The licensee shall verify that the set points for the relief valve on the Accumulators are included in the Inservice Testing Program at the facility. i Response: The Cold Leg Accumulators Relief Valves (NI-52, 63, 74, and 86) are not required to perform a safety function either to shutdown the reactor or to mitigate the consequences of an accident. The inservice testing program requirement to test all class 1, 2, & 3 1 valves was changed to valves which are required for safe shutdown ; of the reactor or mitigating the consequences of an accident.

          ,                                                              .             .i
                                                                                                 )

l I L ! r l 3 Consequently these relief valves are not included in the McGuire Suelear Station pu.:p and valve inservice testing program requ ed l by 10 CFR 50.55a(g). These valves (and setpoints) are tested folicwing maintenance only. j l (Question 13) I l T.S. 3.5.1.2.d 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), revision 39, in Table 6.3.2.1 is necessary to conf 2rm the Safety Analysis Limits for the UHI' Accumulator. It is also- ' i

proposed that it be added as an additional surveillance element to T.S. 4.5.1.2.a. Its absence from the proposed T.S. renders it potentially l non-conservative with respect to the Licensing Basis. The licensee shall j evaluate and propose. l i

The licensee shall verify that the relief valve set point on the Accumulator ! is included in the Inservic'e Testing Program at the facility. ! Response: ! FSAR Table 6.3.2.1 provides the expected operating temperature range j for the Gil accumulator water and not Safety Analysis limits as j stated above. The Safety Analysis value related to UHI water ; temperature is assumed to be the upper bound.value of 100*F. i The Upper Head . Injection Accumulator Relief Valve (NI-279) is not required to perform a safety function either to shutdown the ! reactor or to mitigate the consequences of an accident. The ! Inservice Testing Program requirement to. test all' class 1, 2, & 3 valves was chenged to valves which are required for' safe shutdown of the reactor. or mitigating the consequences of an accident. Consequently this relief valve is not included in the McGuire Nuclear Station pump and valve inservice testing program required by 10CFR 50.55a(g). maintenance only. This valve (and setpoint) is tested following (Question 14) i 11 T.S. 4.5.2.b. Concerning Flow Balance Tests in tne ECCS System. The licensee shall provide the bases for the flow distributions specified and further advise how they might meet minimum flow conditions to intact loops during accident occurrences. Response: The bases for the limits as specified in T.S. 4.5.2.h are the l assumed ECCS flows used in t.b- LOCA analysis. ECCS flow injected j to the broken cold leg is assund to spill in LOCA analyses, so , limits are placed on the branch line totals to ensure that adequate flow reaches the intact loops.

4

(Q-estion 15) T.S. SECT:0N 3/4.5.3 i This T.S. does not disallow the additional CCP and 2 Safety Injection Pumps (SIPS) free 350'T down to 300*. This again is non-conservative with respect te thr: LCOs of the Licensing Basis FSAR which allows only one (1) CCP, and the remaiader i.e., one (1) CCP and any other reciprocating charging pump and 2 SIPS are to be electrically isolated against inadvertent operation. This proposed T.S. is again non-conservative in respect of overpressure protection when compared with the current Licensing Basis. The licensee shall evaluate and propose. The proposed T.S. allows one (1) CCP and one (1) SIP whenever the RCS teep is less than 300'F. The LCO of the Licensing Basis FSAR allows only one (1) CCP because of overpressure 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 licensee shall evaluate and propose. Response: This question appears to be related to the discussion of FSAR Section 5.2.2, "Overpressurization Protection". Although it is stated in two places that Technical Specification 3.5.3.a violates the FSAR Licensing Basis, Section 5.2.2 contains no discussion of ECCS pump operability between 300'F and 350'F. It is further stated, in the discussion of Section 5.2.2. , that the McGuire Technical Specification 3.5.3.a. differs markedly from the Westinghouse Standard Technical Specification 3.5.3.a. Comparing the two we find no differences in the number or type of ECCS pumps required to be operable or inoperable. The McGuire lower limit is l 300*F compared with Standard lower limit of 275'F. We therefore conclude that the McGuire Specification does not differ from the Standard one in a non-conservative manner. (Question 16) T.S. 3.7.1.2.b. l The licensee has deleted operability requirements for the steam-turbine driven auxiliary feedwater pump at steam pressures of less than 900 psig. This is not in accord with current accident analyses and no justification has been provided: Reference 15, Recommendation GL-3, requires the steam-turbine AFW pump in the event of complete loss of AC power for a period of 2 hours and beyond. This will require operability down to the lowest pressures for which 1 the turbine is provided as described in reference 22, Table 10.4.7-6 where the range of operating pressures provided for is from 110 psig to 3205 psig. This will also provide for operability down to and including MODE 4 (and availability from HODE 5) to cover licensing requirements discussed elsewhere under Table 3.3-3, ESTAS INSTRUMENTATION, Items 7a through f. We note two principal features relating to the service conditions of the turbine-driven f eedwater pumps : s

6 ) They are supplied with steam from two steam generators from main steam i lines after Generators. the flow restriction orifices at outlets'from the Steam 3

b. They would normally be expected to perform early in the transient and continue to function according to design flow requirements throughout the occurrence. l The licensee should explain how the proposed T.S. ensures that the turbine driven pump maintains its flow performance required by accident analyses when steam line pressures could drop substantially 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 drift and errors). The licensee shall evaluate the above comments and propose technical specifications which will ensure operability of the turbine-driven AFW pump 4 over the range of conditions expected from design basis accident analysis, and other less boundinE events, down to and including MODE 4 as discussed in the Licensing Basis. Ic his evaluation, the licensee should advise if Item le of Table 3.3-5 l

ESFAS INSTRUMENTATION, Steam Line-Pressure Lov, is ' derived from steam line sensors and after the SG orifices, or if it is taken from pressure sensors on the Steam Generator. The licensee should then advise what has been used in assessing Steam Generator pressure response and turbine driven AFW pump response in the Condition III and especially Condition IV occurrences of the Licensing Basis, and if the existing accident analyses remain valid. Response: The footnote deleting operability requirements for the Steam Turbine-Driven Auxiliary Feedwater Pump (TDAFP) at steam pressures

                                                              <900 psig was added in an attempt to correct a conflict between the LCO with its applicability of Modes 1, 2, and 3 and Surveillance Requirement 4.7.1.2.a.2 which defines operability of the TDAFP as developing a discharge pressure of 1 1210 psig at a flow of 1900 gpm when the secondary steam supply pressure is >900 psig (to delevlop a discharge pressure of 1210 psis the TDAFP requires steam at 1 900 psig, but supply steam pressure can be <900 psig during startups/ shutdowns). The Technical Specification's bases for operability of the Auxiliary Feedwater System is to ensure that the Reactor Coolant System can be cooled down to <350*F from normal operating conditions in the Event of a total loss of offsite power, with the TDAFP capable of delivering a total feedwater flow of 900 GPM at a pressure of 1210 psig to the entrance of the Steam Generators to meet this function. Under normal operating conditions source steam at >900 psig is Available-and the TDAFP is capable of performing this function. However, as indicated in Question 16 and Items 1 and 2 below, the TDAFP is also required with steam pressures <900 psig.

l- ,

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 .       4-t
                   ]

1. During a condition IV feedline break all. steam generators will depressurize prior to closure of the-Main Steamline Isolation Valves (MSIV's). The low steamline pressure set point for closing the HSIV's'is about 585-psig. However, errors due to seismic and environmental conditions as well as instrumentation inaccuracies may result in a stean generator pressure as low as 285 psig prior to MSIV closure. Therefore the turbine driven Auxiliary Feedwater pumps must be capable of delivering-the minimum required flow for feedline break with a steam generator motive supply pressure as low as 285 psig.

2. The ability to commence a plant cooldown must be maintained following transient an'd accident conditions. Following design basis faulted conditions with specific single failure assumptions, it may be'necessary to commence a plant cooldown with only a j

turbine driven Auxiliary Feedwater System pump available. Consequently the turbine driven pump must,be capable of delivering the minimum required flow for.cooldown with a steam generator motive supply pressure as low'as 100 psia corresponding to a primary side hot leg temperature of 350*F during a natural circulation cooldown, which is maximum operating temperature for { Residual Heat Removal System Operation. Therefore, The Tech. Spec's Surveillance requirements / Bases do not adequately define the operability requirements for the TDAFP and ! ! ) consequently the Technical Specification does not ensure operability of the TDAFP over the range of conditions expected from Design Basis Acefdent) Analysis and other less bounding events. All other circumstances (or' accident conditions) besides the limiting condition of loss of Offsite Power during full power operation pose less severe demands on the TDAFP. i for the Main Steamline Break, the intact Steam Generator is fully capable ; of supplying the steam requirements of the pump turbine. With source ' steam < 900 psig the TDAFP is capable of providing feed flow but at a discharge pressure below 1210 psig. Since the McGuire Technical ; Specification is essentially indentical to the Westinghouse Standard. ' Technical Specification (with the exception of the " correcting" footnote), this discrepancy between the LCO and the Surveillance Requirements / Bases should be resolved.on a generic basis and is not specific-to McGuire. , With regard to providing operability down to and including Mode 4 (and availability from Mode 5.), the bases of the auxiliary Feedwater System Technical Specification is that its operability (including the l capacity of the TDAFP) ensures that adequate feedwater flow is available ', to remove decay heat and reduce the Reactor Coolant System Temperature to

            <350*F (i.e. Mode 4) when the RHR System may be placed into operation.

Therefore the bases does not require System Operability in Modes 4 or 5. Since the McGuire and Westinghouse standard technical specifications bases are essentially identical, any desired changes to this bases should be pursued on a generic basis.

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P

l 1 i a Itee le of T.S. Table 3.3-3 " Steam Line Pressure-Low" is derived from l stear lire sensors downstream of the steam generator flow restriction j orifices.. , The steam flow restrictors do not cause a significant pressure ; drop except during a double ended steam line break. The blowdown phase of l this accident lasts only a few seconds. The accurate pressure sensing in i the steam lines (i.e. generation of a " Steam Line Pressure-Low" signal) -( takes less than 2 seconds and steam line isolation less than 7 seconds. ) (The main steam line break accident is descussed in Chapters 6 and 15 of the FSAR). .

                                                                                                      .I (Question 17)

T.S. SECTION 3/4.7.5 I Reference 6, page 9.2-13, revision 39, states that "In the event of solid. layer of ice" forms on the SNSWP, the operating train (of the Nuclear Service ; { Water (NSW} system} is manually aligned to the SNSWP. The Licensee;shall 4 provide the safety-related reason for this action and advise if this t operator action conflicts with the response times proposed under Table 3.3-5. Given a safety Related reason, surveillance requirements ensuring this action l should be included under either T.S. Section 3/4.7.5 NSWS or this particular l T.S. Section 3/4.7.5 STANDBY NSWP. Absent this surveillance requirement on a safety related issue, the proposed T.S. would be non-conservative. The I Licensee shall evaluate and propose. l Response: This action has been deleted. See Section 9.2.2, Nuclear Service Water Syste= and Ultimate Heat Sink, 1984 Update. 1 l (Question 18) T.S. 3/4.9.1 The current SER, Supplement No.1, reference 11, page 15-1, provides that: i

                                                                                                       .1 During refueling the applicant has committed to isolate all sources of unborated water connected to the primary system refueling / canal / spent fuel.

We do note that surveillance requirement T.S. 4.9.1.3 does provide for verifying - that valve no.1NV-250 is closed, under administrative control in support of this. j 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 j current position could be nonconservative if the valve is not specifically locked under the proposed administrative control. Also notice, that reference j 7, page 15.2-14, revision 10, states that. j

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               "The other two paths are through 2 inch lines, one of which leads to the        '

l volume control tank with the other bypassing this tank. There lines contain flow control valves INV-171A and 1NV-175A respectively."

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W are ~.S.s not applied to the closure of these valves also? The proposed

         ~.S. may be nonconservative with respect to the Licensing Basis. The licensee shall evaluate and propose.

i Respense: Valve IVV-250 is specifically required to be locked closed under l the Administrative Controls (i.e. Station Procedures). This Valve is upstream of valves 1NV-171A and 1NV-175A and isolates the flow path. 1 (Question 19)

{ T.S. SECTION 3/e.9.8 The ACTION statement provides that with no RHR loop operable, the containment should be closed within 4 hours. Information in reference 8, page Q 212-56 under Case 2 shews that if RHR is absent [by isolation of the RCS/RHR inlet valve] that:

              "Approximately 2.5 hours are available to the operator to establish an alternate means of core cooling. This is the time it would take to heat 300,000 gallons of water in the refueling canal from 140'T to 212'F, assuming the maximum 24 hours decay beat load."

The current value of 4 hours appears less conservative than this calculated value of 2\ bours in the FSAR. The licensee shall evaluate and propose. Review of available responses to the consequences of a fail closed RCS/RHR isolation valve, include many procedures using the containment sump. To allow for this single failure contingency, the licensee should therefore ensure that the contain=ent sump will be operable during this mode, and with an appropriate 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. The licensee shall evaluate and propose. Response: The McGuire Technical Specification 3.9.8 is the same as the Westinghouse Standard Technical Specification (STS) 3.9.8. Since there is nothing unique about McGuire's 3411 MWt power level, its decay heat characteristics, or its 23 feet level requirement, this question should be addressed on a generic basis. (Question 20) T.S. SECTION 4.9.8.2 The current ACTION statement calls for containment closure in 4 hours (i.e. 240 mins). Earlier conservative calculations for this MODE show that loss of all RRR in this MODE can cause boiling in 5 minutes and core uncovery in 100 mins. Given the circumstances, containment enclosure should be effected

o I l l traeciately, coemencing RFJi low flow alarms. The Licensee shall evaluate, and propose. The current T.S. appears nottonservative with respect to the Lice:. sing Basis, , j hespcase: See the response te the previous item since McGuire is also in accordance with Westinghouse Standard Technical Specification on t r.i s 'te=. 1 i l l l l 1 I 1 r 1 1

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Dear Mr. Chairman:

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4.1 CONCLUSION

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