ML20126D178
| ML20126D178 | |
| Person / Time | |
|---|---|
| Site: | McGuire, Mcguire  |
| Issue date: | 06/03/1985 |
| From: | Licciardo R Office of Nuclear Reactor Regulation |
| To: | Harold Denton Office of Nuclear Reactor Regulation |
| Shared Package | |
| ML20126D181 | List: |
| References | |
| NUDOCS 8506140668 | |
| Download: ML20126D178 (2) | |
Text
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W N 8 3 Igac MEMORANDUM FOR:
Harold R. Denton, Director Office of Nuclear Reactor Regulation FROM:
Robert B. A. Licciardo, Nuclear Engineer Section A, Reactor Systems Branch, DSI
SUBJECT:
MCGUIRE TECHNICAL SPECIFICATIONS:
REQUEST FOR M TION OF POST RESOLUTION PHASE OF DIFFERING PROFESSIONAL OPINION OF MR. R. LICCIARDO The writer has recently become aware, that only 220 items of the 380 concerns identified by him during the Post Resolution phase of his subject DP0 have been forwarded by RSB to DL for Licensing Action.
Further, the writer has been informed that the remaining 160 items have been closed out by RSB alone, and without the necessary Lice' sing Actions required by 10 CFR for this now Operating Reactor Facility of 2X3411 megawatt thermal Units which has already been in operation since 1981 for McGuire Unit 1, and 1984 for McGuire Unit 2.
Related correspondence in this matter is provided under Enclosures 1 and 2.
Consequently, in conformance with the Agreement and Commitments in this matte' deriving from the earlier Resolustion of his DP0, as described in Enclosures 3, 4 and 5, the writer asks that you appoint others who are able and willing to complete this task.
To facilitate complete and final necessary Regulatory Action for this Facility i
I recommend:
a)
Completion of the Licensing Action by DL on the limited set of 220 items currently under review.
b)
In parallel with a) above, the Licensee be informed of a following complementary set to ensure a Complete and Safe Licensing Action within the current schedule for Implementation, c)
Completion of the remaining 160 items by Division of Licensing.
Since DL was asked by RSB/DSI to review their Categorization of the initial set of 220 items and since they presumably have already established a basis for defining the necessary related Licensing Actions, it is prudent to enable treatment of the remaining 160 items in a similar and thereby consistent manner. to this memo is a marked version of the writers " Review of McGuire Technical Specifications" identifying the referenced 160 items.
The remainder of this document was used directly by RSB/DSI in its transmittal to DL for 8506140660 850603 PDR ADOCK 05000369 P
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H. Denton -
Licensing Action, and so the interrelationships of the remaining items with the existing activity are clearly defined for the purpose of the required Completion.
I await Confirmation of your intent to complete this review, and finally place this Facility in a Completed Regulatory status and also to restore the necessary assurance of the Level of Public Health and Safety intended in its original Licensing basis and as subsequently Amended.
Yours truly, Ro N d Y N N Iardo B. Mech. E; B. Comm.
Professional Nuclear Engineer:
No. NU 001056 (California)
Professional Mechanical Engineer:
No. M015380 (California)
Enclosures 1.
Memo, April 24, 1985:
R B. A. Licciardo to B. W. Sheron.
2.
Memo, May 8, 1985:
B. W. Sheron to R. B. A. Licciardo.
3.
Memo, February 27, 1984:
L. H. Barrett to H. R. Denton.
4.
Memo, March 20, 1984:
H. R. Denton to R. B. A. Licciardo.
5.
Memo, March 21, 1984:
H. R. Denton to D. G. Eisenhut and R. J. Mattson.
6.
Memo, June 11, 1984:
R. B. A. Licciardo to B. W. Sheron w/ attachment entitled:
McGuire Units 1 and 2 Proposed Technical Specifications Review Of " Proof and Review Copy Prepared by Robert B. A. Licciardo, June 12, 1984.
cc:
Nunzio J. Palladino, Chairman w/ attachment Commissioner Roberts DISTRIBUTION Commissioner Asselstine Docket File Commissioner Bernthal RSB R/F.
Commissioner Zech RSB P/F: McGuire Dockets SECY RLicciardo/McGuire DP0 File OPE w/ attachment OGA PDR:McGuire Dockets 50-369/370 CA w/ attachment W.J. Dircks w/ attachment RLicciardo R/F H. Denton w/ attachment RLicciardo D. Eisenhut DOCUMENT NAME:
RL 11, KAL H.L. Thompson w/ attachment T. Novak R.M. Bernero w/ attachment W. Houston B. Sheron N. Lauben "0FFICIAL RECORD COPY" R. Licciardo w/ attachment D. Hood w/ attachment DSI:RSB RLScciardV.jf Wica/85
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ENCLOSURE 1 APR y y MEMORANDUM FOR:
Brian WT Sheron, Chief 7-Reactor Systems Branch 1,.
Division of Systems Integration THRU:
G. Norman Lauben, Section Leader Section A Reactor Systems Branch Division of Systems Integration FROM:
Robert B. A. Licciardo, Nuclear Engineer
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Reactor Systems Branch Division of Systems Integration
SUBJECT:
MCGUIRE TECHNICAL SPECIFICATIONS
Reference:
a) Memorandum, Dirks 'to"Palladino "McGuire Technical Specifisations" dated April 12, 1985 b) Memorandum, Bernero 4.o Eisenhut, concerns on McGuire
. Technical Specifications dated August 30, 1984 The writer has recently received a "For Your Information" copy of the'1etter to reference a) and in anticipation of an early close out of this issue,
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briefly reviewed t.he completeness,of the activity by RSB and DL..The review,
reveals that of approximately 380 items submitted by the writer, only 220 are
(,beingreviewe'dbyDL.
The lateness of the observation is caused by the language of refbnce b) which
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the writer interpreted as meaning that all concerns were to be reviewed by DL with.an initial Set having air.eady been,categoriz,ed by RSB.
The writer must conclude that the review of the remaining 160 items, by DL, is necessary to ensure a valid, safe, and complete action.
From'its review, the RSB staff appear to have selected a Set of Technical Specifications [TSs), for priority action, which are primary reflections of a number of Principal Iss*ues of Concern.
However each such Concern has a number of complementary T.S. requirements which must be changed or added, to ensure a I
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complete valid protection.
For example:
A modification to increase the number of Reactor Coolant Pumps in Modes 3, 4 [and 5] must be complemented by
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Actions in a number of other TSs, such as " Control Rod Insertion Limits",
which have been excluded.
Other items have their basis for inclusion in Regulatory Req 0irements, and should be included for consideration by the same set of principle ~s with which the priority material is being adjudged.
At this time [I would recommend completion of the current action'on the exist-ing priority items.
In parallel with that, the licensee should be informed of a following complementary set, to ensure a complete and safe licensing action within the current schedule for implementation.
~ An early decision on this action is necessary to enable DL to incorporate related elements and conditions in its proposed implementation with the licensee scheduled to commence May 1, 1985.
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Robert B. A. Licciardo, Nuclear Engineer Reactor Systems Branch Division of Systems Integration cci R. Bernero R. Houston DISTRIBUTION N. Lauben..
Docket File ~
RSB R/F RLicciardo R/F NLauben BSheron RLicciardo MCGUIRE SPEC KAL
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ENCLOSURE 2
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MEMORANDUM FOR:
Robert Licciardo, Reactor Systems Branch, DSI FROM:
Brian Sheron, Chief, Reactor Systems Branch, DSI
SUBJECT:
MCGUIRE TECHNICAL SPECIFICATIONS
Reference:
(1) Memorandum, Licciardo to Sharon, "McGuire Technical Specifications," dated April 24, 1985 (2) Memorandum, Bernero to Eisenhut, " Concerns on McGuire Technical Spe::ifications," dated August 30, 1984 e
(3) Memorandum, Sheron to Licciardo, "McGuire Tech Spec Assignment,". dated April 11, 1985 I am writing in response to your reference (1) memorandum in 'which 'you observe that, due to the language of reference (2), you interpreted it to mean that all of your concerns would be reviewed by DL, rather,than the subset that resulted from the RSB management categorization.
You also conclude in your reference (1) memorandum' that review of the remaining 160 items by DL is "necessary to ensure a valid, safe, and complete action."
In~ response to the first item, I believe that the language in reference (1) was cTear and self explanatory regarding.which of your items would be for-warded to DL and which ones wouldn't.
A copy of the cover lettet of refer- -
enge (2) is provided as Enclosure (1).
Regarding your second item, RSB management spent a considera61e amount of time and resources reviewing your approximately 380 concerns.
Notwithstanding this latest expression of your desire to have DL review the concerns for which we found no merit, I must advise that our previous review of your work and our conclusions stand as is.
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FJ Brian Sheron, Chief Reactor Systems Branch Division of Systems Integration
Enclosure:
As Stated cc:
H. Denton D. Eisenhut R. Bernero R. W. Houston G. N. Lauben G
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AUGS0W MEMORANDUM FOR:
Darrell G. Eisenhut, Director Division of LTeensing FROM:
Robert M. Bernero, Director Division of Systems j
Integration
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SUBJECT:
CONCERNS ON MCGUIRE TECHNICAL.SPECIFICATI0KS
Reference:
Memorandum, Sheron to Denton, '" Review Status 'of Technical Issues on McGuire Tech Specs" dated June 25, 1984 In the reference memorandum, Mr. Denton was advised that the RSS mge-ment would review the concerns of R. Licciardo on the McGuire technical specifications as he clarified them in his June 11, 1984 memorandum and forward the results to DL.
RSB has completed its review and categoriza-tion of the concerns, and this memorandum forwards the results to your office for disposition.
In summary, we identified. no concerns of sa ty significance that re-quired immediate action, and all concerns cauld be addressed.as part of the process described later on in this memo.
Our categorization process eliminated those concerns that RSB management felt were either not appropriate for technical specifications or still did not clearly specify the issue.
The remaining concerns were catego-rized as either category A, tho.e concerns that were plant specific within the scope of the standaro Technical Specifications and were-appropriate to ark an applicant, and category B, concerns that were felt to be philosophic in nature, questioning the scope and content of the technical specifications.
The category A concerns are provided in enclosure (1) and the category B concerns are provided in enclosure (2).
is With regard to the category A items, these are questions which the RSB management felt were appropriate to be asked of an applicant, but not necessarily considered to be final " positions."
Based on the response,,
the staff would have to decide whether it was acceptable or if changes to the McGuire and standard technical specifications were warranted.
If it were the latter, we would follow the Office Letter 38 guidance.'
We also note that the categorization process was donk by 5 managers.
Different judgments could result in some differences in categorization.
.You should therefore feel free to recategorize those items you believe are.miscategorized.
We have worked with Cecil Thomas of your staff and have agreed orn the following approach to final resolution:
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DL will review the category A cnd B. items tnd identify 1.
those for which they believe acceptable answers already exist for,the Technical Specifications.
These concerns and the answers,will be documented by DL.
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2.
Of the remaining concerns, DL will review the categoriza-tion and revise them as necessary into items which are
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plant specific to McGuire, items which are generic, and items which are applicable to both.
~ 3.
For those items that are generic, they will be returned' to DSI by DL for consideration by DSI for incorporation in
, the.next periodic update of the standard technical speci-fications--in accordance with the provisions of Office Letter 38.~
4.
For those items that are plant specific, DL will determine I
how to addres's.them with the McGuire licensee.
L DSI (RSB) will assist DL as nec'essary in carrying out these final steps of the resolution plan.
kisinalSisnedf.I, is..rsu.s.rnere I
Robert M. Bernero, Director Division of Systems Integration
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Enclosure:
As stated i
DISTRIBllTION i
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.H. Denton Central Files l
E. Case RSB R/F O. Crutchfield RSB S/F: Licciardd DP0 C. Thomas BSheron R/F F. Miraglia AD/RS Rdg.
D. Brinkman RBernero R. Birkel BSheron T. Novak E. Adensas RSB S/L's i
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February 27, 1984 9,
.y MEPORANDUM FOR:
Harold R. Denton, Direttor Office of Nuclear Reactor Regulation FROM:
Lake H. Barrett, Deputy Program Director TMI Program Office t
SUBJECT:
REC 0P91 ENDED RESOLUTION OF R. LICCIARDO DP0 As requested by your memo of December 29, 1983, I have conducted an independent assessment of the December 7,1983 R. Licciardo DP0 (Enclosure 1) f concerning disparities between the McGuire reactor safety system technical
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specifications and the safety analyses of record in the licensing documents.
Mr. Licciardo provided further description and elaboration on his DP0 in memoranda to me dated January 26 and 27,1984 (Enclosures 2 and 3). At sqy
. request, during the month of February, Reactor Systems Branch management spent.
many hours reviewing with Mr. Licciardo the technical substance and bases of
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his specific concerns. DSI supplied infomation retgarding this DP0.is "
attached as Enclosures 4 through 9.
s I have evaluated the documents and met with the various parties and have concluded that the issue raised in Mr. Licciardo's DP0 warrants further staff attention.
I recomend the following actions:
1)
In accordance with NRC Manual Appendix 4125, Section G.1.a. adopt the views of Mr. Licciardo's December 7,1983 DPO. This OPO addresses apparent disparities between the McGuire reactor safety system technical specifications and the safety analyses of record within the licensing. documents.
2)
Develop and implement a plan' for timely identification and resolution of the McGuire disparities.
3)
Perform a review of staff procedures and practices used for the review of technical specifications when issuing operating reactor licenses.
It is my understanding the DL presently has such an effort underway.
It is difficult to assess the safety significance of this disparity issue before a more complete technical review of the McGuire disparities is completed. Based on my discussions with Mr. Licciardo and other staff members' I consider this issue important deserving staff attention. As Mr. Licciardo
. states in his DP0 the disparities "suggest" that regulations "cculd be I
compromisedd and that compromises "could manifest" in increased risk. Sty limited' review of Mr. Licciardo's elaboration of the disparities in his 0.
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.r Harold R. Denton 2
Febru ry 27, 1984
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January 26, 1983 memorandum (Enclosure 2) indicates that the disparities in the McGuire technical specifications would not reduce overall safety s:argins to a point resulting in an unacceptable public r'isk and,' therefore at 'this time, do not require extraordinary regulatory actions. Once the staff has completed its safety review of the disparities appropriate regulatory action can then be determined.
Mr. Licciardo's December 7,1983 DP0 requested a impartial peer group review in accordanco to Manual Chapter 4125.
I have since discussed my above.
proposed resolution with Mr. Licciardo and he has agreed to waive the peer group review provided the above resolution is adopted.
A Lake H. Barrett Deputy Program Director i
Tl;I."rogram Office
Enclosures:
1.
Memo, December 7,1983, R. B. A. Licciardo to G. Norman Lauben 2.
Memo, January 26, 1984, Robert B. A. Licciardo to Lake H. Barrett 3.
Memo, January 27, 1984, Robert B. A. Licciardo to take H. Barrett 4.
Memo, December 9,1983 Brian W. Sheron to Robert Licciardo
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5.
Memo, December 13, 1983, G. Norman Lauben to Brian W. Sheron 6.
Memo, December 15, 1983, Brian W. Sheron to Harold R. Denton 7.
Memo, December 15, 1984, R. Wayne Houston to Roger J. Matt, i 8.
Memo, February 1, 1984, Brian W. Sheron to Lake H. Barrett 9.
Memo, February 22, 1984, Brian W. Sheron to 4 eke H. Barrett cc w/ enclosures:
R. Licciardo R. Mattson D. Eisenhut J. Carter B. Sheron '
F. Miraglia C. Thomas D. Brinkman R. Houston N,. Lauben E. Adensam R. Birkel R. Majors
.Licciardo DP0 File e
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O ENCLOSURE 4
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March 2D,1984 MEMORANDUM FOR: R.B.*A.'LicciaNi@
' Reactor Systems Branch ,
D'ivision of Systems Integration FROM:
Harold R. Denton, Director Office of Nuclear Reactor Regulation
SUBJECT:
YOUR DIFFERING PROFESSIONAL OPINION REGARDING DISPARITIES IN THE MCGUIRE TECHNICAL SPECIFICATIONS You raised issues concerning disparities ' involving technical specifications at McGuire.
These disparities have been evaluated and on the basis of recommendations presented by Mr. Lake Barrett in a February 27,1984.
. memorandum to me, I have initiated certain actions.
In accordance with Manual Chapter 4125, Differing Professional Opinions, enclosed is a copy of f-my memorandum which provides a description of the actions being taken to
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resolve technical issues expressed in your differing professional opinion.
/
Harold R. Denton, Director Office of Nuclear Reactor Regulation
Enclosure:
As stated cc:
R. Mattson D. Eisenhut J. Carter B. Sharon F. Miraglia C. Thomas D. Brinkman i
R. W. Hopston sy
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ENCLOSURE 5 gd:ssg' %
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March 21,1984 3
MEMORANDUM FOR:
Darrell G. Eisenhut_, Director Division of Licensing Roger J. Mattson, Director Division of Systems Integration FROM:
Harold R. Denton, Director Office of Nuclear. Reactor Regulation
SUBJECT:
DIFFERING PROFESSIONAL:0 PINION OF MR. LICCIARDO REGARDING MCGUIRE: TECHNICAL SPECIFICAT10N 3
Mr. Licciardo, NRR, on December 7,' 1983 submitted a Dif'fering Professional-Opinion (DPO) concerning ~ disparities between the McGuire technical.
specifications and the staff safety evaluation.
I subsequently gave Mr. Lake Barrett the assignment of assessing the DPO.
Mr. Barrett 'provided me his
- assessment and recommendations in the enclosed. memorandum dated February 27, 1984 I have evaluated his' assessment.and have decided.to pur. sue further -the evaluation of specific disparities.at.McGuire and' the adequacy of procedures used by the staff when developipg the technical specifications required for
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facility operation.
The Division of Licensing shall review the. adequacy of' staff procedures and
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the actual practice used in development of technical specifications for an operating license.
Existing procedu.res shall be modified. if appropriate, and a brief report sent to'me that summarizes the review and conclusions..
The report on your effort should be completed no later than.May 1,1984 The Division of Systems Integratio.n, 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.
As soon as the review approach is sele'eted, you are to provide me l
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 than July 1,1984 or provide a status report with a schedule by May 15, 1984 Pursuant to the procedures for resolving a Differing P'rofessional Opinion, Manual Chapter 4125, I consider the DP0 resolved.
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Marold R. Centon, Director Office of Nuclear Reactor Regulation Enclem:
As stated cc:
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R. Licciardo J. Carter B. Sheron F. Pliraglia C. Thomas D. Brinkman R. W. Houston t
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O ENCLOSURE'6 This is a marked version of the " Review of McGuire Technical Specifications" identifying the 160 items excluded from initial consideration, by a vertical bar in the right-hand margin.
The remainder of this document was used directly by RSB/DSI in its transmittal to DL for Licensing Action and so the interrelationships with the existing activity are clearly defined for the purpose of the required Completion.
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MEMORANDUM FOR:
Brian W. Sheron, Chief Reactor Systems Branen
. Division of Systems Integration FROM:
Robert B. A. Licciardo Nuclear Engineer Reactor Systems Branch Division of Systems Integration
SUBJECT:
REVIEW OF.MCGU, IRE TECHNICAL SPECIFICATIONS
REFERENCE:
a) Memo from Harold R. Denton, Director Office of Nuclear Reactor Regulation t
for Darrell G. Eisenhut, Director i
Division of Licensing and Roger J. Mattson, Director Division of Systems Integration on the
Subject:
DIFFERING PROFESSIONAL OPINION OF MR. LICCIARDO REGARDING MCGUIRE TECHNICAL SPECIFICATION and dated:
March 21, 0.984 b) Memo from Brian W. Sheron, Chief RSB DSI to Robert Licciarce RSB, DSI dated April n,1984 on the
Subject:
MCGUIRE TECHNICAL SPECIFICATIONS ASSIGNMENT
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l I reference your memo to reference D) requesting review of the McGuire Technical Specifications to an acceptacle fermat, in eesponse to the requirement of reference a) for a coordinated review of the concerns arising from the writer's earlier DPO.
Please find attached copy of a dccument entitled "McGuire Units 1 & 2:
Proposed Technical Specifications; Review of Proof and Review Copy," wnich is in response to your request.
The review is com;:osed of two sections.
The first section is entitled " Pre Review Information" wnien cetails the Basis, Purgese anc, Resources, Schecule, Evaluation Metned, Regulatory Requirements and Licensing Consecuences of the Review.
The second section contains the Detailed Review.
Since. the staff recuired this detailed review te be conducted without any formal, or substantive infermal ciscussion, botn within and without RSS' I
= resume 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 Integratien, in ecercinatien with DL, sna11 have people that are katowledgeable about the technical su:jects raised y Mr. Licciardo, the standard tecnnical specifications, and the M:Gwire technical specifications
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17 writer considers that such a coordinated review including constructive critique is an essential consequence of any such cocument.
The writer also believes that such construction must be developed on the basis of responsible written and signed comment within the Regulatory Framework.
The writer would be pleased to participate in this coordination as required.
The writer is aware that RSS staff has received copies of the writer's initial proposed memo to T. M. Novak from R. W. Houston on the subject of:
"$TAFF REVIEW 0F PROOF AND REVIEW COPY OF PROPOSED TECHNICAL SPECIFICATIONS MCGUIRE UNITS 1 & 2" dated 06/15/83, and through this action is pleased to have made an early contribution to recent reviews of Technical Specifications for Operating License Applicctions.
Further, the writer has been informed that the above referenced memo (of 06/15/83) was also provided to Westinghouse (y) and notes two subsequent developments of significanca:
1)
In response to a question from M. Wigdor concerning "Vogtle," on " Cold Overpressure Mitigation", W has now recently submitted a Topical report entitled " Cold overpressure Mitigating Systems," dated February 1984, for review by NRC.
2)
W has recently reviewed its position on Reactor Coolant System (RCS) 5perability requirements in MODE 3 anc from this has determined the neog for acettional opersole RCS pumps over those required in the W STS for the casa cf " Uncontrolled Rod Cluster Control Assembly Bank Withdrawal
,From a Sumeritical Concition."
Both of the above items 1) and 2) were the$ subject of specific concern in the referenced memo proposed by the writer, and it is encouraging to note the early resconse by y to those safety issues.
WY R. B. A. Licciardo
Attachment:
As stated O!STRIBUTION Central N1e cc:
H.R. Denton RSE R/F R. Mattson RLicciardo R/
RLicciardo DP0 Nie R. W.
Houston w/ attachment RLicciardo N.
Lauben w/ attachment a e :i, sa i G**ic e >
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,, * '. 3 MCGUIRE UNITS 1 & 2:
PROPOSED TECHNICAL $PECIFICATIONS REVIEW OF " PROOF & REVIEW COPY" Prepared By ROBERT 8. A. LICCIARDO Nuclear Engineer RS8/DSI/RS8 Date: June 12,1984 O
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O,A A a TA8LE OF CONTENTS PRE REVIEW INFORMATION BASIS OF REVIEW PURPOSE OF REVIEW SCHEDULE AND RESOURCES EVALUATION METHOD REGULATORY REQUIREMENTS LICENSING CONSEQUENCES OF REVIEW INVITATION FOR COMMENT OETAILED REVIEW A00ENOA:
Later items For Consideration List of References Table 1.
Sections Reviewed 8y Reactor Systems Branch Table 2.
Technical Specification Paces Affected APPENDIX A:
Technical Specifications - Selected Relevant Reautations iii Revision A
l INTRODUCTION By letter to reference 1), the licensee proposed Technical Specifications for 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 use for Unit 2 but which would enable its use for Unit I at a later date.
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 appifcation for a proposed, and issuance of a subsequent, licensing amendment at a later datt.
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/0L and SSPB, subsequent changes had been made.
This subsequent document presented to R$8 for review, contained no record of, or, safety evaluation reports on, these changes which had been made including any relationship to the then existing McGuire Unit 1 Technical Specification and the Final Safety Analysis Reports, or the Safety Evaluation Reports, 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 items reviewed are listed in Table 1 and the pages affected are listed in Tacle 2.
l 06/01/S4 111 Revision A L
o.
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 Proof 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 (SSP 8) 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 serve the purpose of the request by Harold R. Denton in Reference a) namely:
"The Divison of Systems Integration, in coordination with OL, 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 OPO."
~
For this purpose, RS8 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 of a set of information within a time frame con-sistent with the proposed purpose and schedule, the writer was asked by RS8 to complete his task with minimal interchange both within and without RSS.
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 2
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 Mar.h 21, 1984, Harold R. Denton required that:
"The Division of Systems Integration, in coordination with OL, shall 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
l e
review the broad technical subjects and subgroups raised in the OPO.
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 than July 1,1984 or provide a status report with a schedule by May 15, 1984."
I i
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 l
the probabilities that because of the relatively simple nature of the related 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 many locations of lesser perceived importance incl.uding. footnotes, and descrip-tions in the Basis, att, ached.to the T.S.
These co.ncerns have. required a near i
it'em 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 1
responsibility to ensure the accuracy and completeness of the Technical Speci-fications including a final detailed check and evaluation of not only the items that are covered above, but residuals in the area of unconfirmed concern for RSB.
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 l
- 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.
l The term " evaluate," used throughout this review as e.g., in the phrase "The 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 " propose" 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 $50.36" and 10 CFR S50.59 " Changes, tests and experiments" in respect of " proposed change, test or experiment."
Regulatory Recuirements 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 I
" Technical Specifications." This defines the principal Require-ments which will be included in the Technical Specifications.
These include:
" Safety limits, limiting safety system settings and limiting control settings."
" Limiting conditions for operation" 10 CFR 50.36(c)(3)
" Surveillance requirements" 10 CFR 50.36(c)(4)
" Design Features" 10 CFR 50.36(c)(5)
"Adminstrative. controls" 10 CFR 50.11
" Exceptions.and Exemptions from Licensing Requirements",.
" Specific Exemptions" These two Regulations define the basis for granting exemptions from the requirements of 10 CFR.
10 CFR 50.34
" Contents of Applications:
Technical Information" This pr~ovides the regulatory basis for a)~ Necessary descriptions of the facility and the need for 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 determined as the result of preliminary safety analysis and evaluation to be probable subjects of technical specifications for the facility, with special attention 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 10 CFR 50.34(b)(6)(vi) 10 CFR 50.57
" Issuance of Operating License" The particular relevant subsections are:
10 CFR 50.57(a)(1) - This ensures that the facility has been substantially constr'4:ted, 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...
3 Revision A
s.
"Each operating license will. include appro-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 such items will not endanger public health and safety."
" Changes, Tests and Experiments" 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 unreviewed 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 Safety Evaluation, whether involving 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.
" Application for amendment of license or construction permit" This provides 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 acoli-cable the form prescribed for original aaplications."
10 CFR 50.100 " Revocation, suspension, modification of licenses and construc-tion permits for cause."
Licensing Consecuences of Review The consequences of the review in terms of the types of problems encountered in meeting regulatory requirements may be categorized as follows:
1)
Descriptions which are incomplete, ambiguous and errored, varying from relatively minor matters to matters of substantial importance to =afety.
Except for relatively minor matters, this category has been considered non conservative since they provide no sound basis for ensuring that the detailed requirements of the Licensing Basis are specified for the operating facility.
4 Revision A
,o 2)
Plant Engineering providing for unlimited operability of Process and 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 Basis 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 Licensir.g 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 to Regulations pertaining to that Branch.
Examples include proposed deletion of auto initiation of MD-AFW pumps below P-ll by manual block, and deletion of Pressurizer Water Level - Hign trip.
The proposed 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 procose.
This circumstance also introduces mixed and deficient protection rationale for a large number of occurrences re' quiring protection under Regulatory Requirements.
4)
Plant Engineering with Operability limited by Plant Control Logic.
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-analy:ed events.
Examples include the absence of any " Low Flow" Reactor Trips below the P-7 permissive, and absence of many nther Reactor Trips.
5 Revision A
I a.
The plant is inside the Licensing Basis Engineering which however has not been adequately evaluated. This is a situation in which Requiatory Requirements have not been met within the ensuing Licensing Basis since an adequate clarification of and evaluation of the circumstances has not been undertaken.
The ifcensee 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 i
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 Boration Control in MODES 1, 2, 3 and 4 and Emergency Core Cooling Systems in MODES 3, 4, and 5.
No evaluation and proposals are submitted.
r The Licensee shall evaluate and propose.
I 7) b)
Also, as a result of 7)a), we have discussed possible modifications l
to these proposed Technical Specifications, which may make them acceptable providing appropriate protections 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 MC'IS 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 reouirements i
for equipment performance and that performance necessary to achieve the required Plant Protection under Condition II, III and IV Occurrences.
l l
6 Revision A I
,o 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 ultimately 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 enormity 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 Documents.
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.
Invitation for Comment The writer would welcome written and signed comments within the Regulatory Framework, on this Review.
Re f e aene es
~
a)
Memo from Harold R.
Denton, Director Office of Nuclear Reactor Regulation for DarreLL G.
Eisenhut, Director Division of Licensing and Roger J.
Mattson, Director
~~
Division of Systems Integration on tne
Subject:
DIFFERING PROFESSIQNAL OPINION OF MR. LICCIARDO REGARDIt!G MCGUIRE TECHNICAL SPECIFICATION and dated:
March 21, 1984 b)
Memo from Brian W.
Sherone Chief, RSS, DSI to Robert Licciardo RS8e DSI dated April 11, 1984 en the
Subject:
MCGUIRE TECHNICAL SPECIFICATIONS ASSIGNMENT 7
Revision A
MCGUIRE UNITS 1 & 2:
PROPOSED TECHNICAL SPECIFICATIONS DETAILED REVIEW OF " PROOF & REVIEW" COPY PREPARED BY Robert 8. A. Licciardo Nuclear Engineer RS8/OSI/RSRS Date:
June 12. 1984 06/07/84 Revision A
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,yg) shall not exceed the limits shown in Figures 2.1-1 and 2.1-2 for four and three loop operation, respectively.
APPLICA8ILITY: MODES 1 and 2.
ACTION:
Whenever~the point, defined by the combination of the highest operating loop average temperature and THERMAL POWER has exceeded the appropriate pressurizer pressure line, be in HOT STAN08Y within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />, and comply with the requirements of Specification 6.7.1."
EVALUATION a)
Concerning the title:
SAFETY LIMITS / REACTOR CORE.
Clarify if the numerical values in Figure 2.1 are meant to be Safety Limits, Limiting Safety Settings or Set Points.
b)
Concerning Figs 2.1-1 What is the licensing basis for this type of re-presentation, i.e., RCS T,yg (*F) vs Fraction of Rated Thermal Power, and the values in this figure.
Reference 7, Figure 15.1.1-1, revision 7 is the existing Ifcensing basis; it provides different ordinates, T,yg vs aT and includes descriptions of related acceptance criteria and limits which 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; :nis 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.
06/01/84 1
Revision A
t, e)
Concerning description under Section 2.1.1 above. We propose this de-scription should clarify that the " combinations" presented are those allowed under " Anticipated Operational Occurrences" and not steady state conditions.
f)
The FSAR does describe a constrained set of thermal hyr.*aulic parameters for the Reactor Coolant System under steady state normal operating con-ditions upon which " plant safety" under Condition II, III and IV Occur-rences is established.
These are generally described in reference 7, under Section 15.1.2, Table 15.1.2-2, and the programmed T,yg provided under reference 3, Figure 5.3.3-1; pressurizer pressure is provided under Table 5.1-1.
(Related pressurizer level and steam generator levels will be discussed under T.S. Sections 3/4.4.3 and 3/4.4.5) Should not these values be included in the Technical Specifications (in appropriate set point methodology) to meet the requirements of 10 CFR 50.36.
For the thermal-hydraulic parameters represented in Section 2, the steady state set points would be represented by a single line showing programmed Tavg against programmed.1T 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
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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.
ApPLICABILIp_: MODES 1, 2, 3, 4, and 5.
ACTICN:
MODES 1 and 2 Whenever.the Reactor Coolant System pressure has exceeded 2735 psig, be in HOT STANOBY with the Reactor Coolant System pressure within its limit within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />, and comply with the requirements of Specification 6.7.1, MODES 3, 4 and 5 Whenever the Reactor Coolant System pressure has exceeded 2735 peig, 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 pr.essurizer pressure shall not exceed the 06/01/84 2
Revision A
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 ender any Anticipated Operational Occurrence or Design Basis Accident.
I 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 pressurn " What provision has been made in the specified value or related instt.amentation to conservatively
' account for this necessary co.rrection, d)
Please clarify that the value'of 2735 psig is an act6a1 Safety Limit, being 110*. of the Design Pressure of 2485 psig (reference 3, Table 5.2.2-2) and how is such a value determined by the operator when no set point, allowable values and channel errors are provided for or defined, e)
Concernihg jer, ion Statement: MODES 1 & 2.
This should consider restora-tion of the iUS 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 generally identical concerns prevail except for the limited Applicability of Appenilix G in T.S. Figs. 3.4-2.
f)
Concerrging f900ES 3, 4 & 5.
How is the pressure limit of 2735 psig applicable to MODES 4 and 5 when 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 Appendix G limits of <2500 psig at RCS temcs, 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.
l l
l 06/01/84 3
Revision A
~
i L
TABLE 2.2-1.
REACTOR TRIP INSTRUMENTATION SET POINTS 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 described as applicable to McGuire Unit 1, 50-369.
At this date, the assump-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 g a certain value is less than satisfactory; it is open-ended allowing overly conservative setpoints with unnecessary reactor trips.
It appears that the Set-Point methodology may already have provided for expected errors in setting SETPOINTS so that this open-ended uncertainty is eliminated to a satisfactory
" manageable" quantity.
The Licensee should clarify.
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.
Atem 4.
Power Rate, Neutron Flux, High Negative Rate.
Will a time constant of >2 seconds result in a slower response time which is less conservative?
Reference 18 page 3-13, concerning Set Point Methodology advises that this value is not used in Safety Analyses.
This appears in direct contradiction to reference 7, Section 15.2.3, page 15.2-12, revision 7, first para.
The Licensee shall evaluate and propose ftem 5:
TS incomplete; should read as:
Intermediate Range, (High] neutron flux.
Item 9:
Pressurizer Pressure-Low The specified. Trip Setpoint & Allowable values agree with those provided uncer 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 referenco 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 During Power Operation",
page 15.2-40, revision 43 item 7, " Reactor Trip ----- is initiated by low pressure at 1800 psia;" This is however relatively conservative with respect to the other values used above.]
The Licensee shall review and clarify.
Etem 17:
The existing descriptor " Safety Injection Input from ESF" should be replaced by " Reactor Trip from ESFAS."
06/01/84 4
Revision A
The fo!1owing 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 1e under frequency, undervoltage, loss of RCP breaker.
Such anticipatory sig.nals are provided below P-8 when, safety consequences are more conservative for.this f'cility.
(See later 12b.)
Is this adequate conformance to diversify require-a 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 107..
For conservatism, should the Safety Analysis Lf;.a.t be the W STS value less 10*.'; is this neces-sarily conservative for all Licensing Basis occurrences.
Item 14: When two or'more RCP circuit breakers open, above Permissive 7 (10P.
power), Reactor Trip deriving from undervoltage of the Reactor Coolant Pumos is also initiated, reference 7 Section 15.2.5.1 and reference 5, figure 7.2.1-1 06/01/84 5
Revision A
note 4.
It is proposed that a notation to this effect should appear under this item.
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.
i Item 18b:
Low Power Reactor Trips Block, P-7 i
a)
This T.S. provides that when power level is less then Permissive P7 (with P10 (Nuclear) or P13 (turbine) powers of less than 10%) the undervoltage (and RCP breaker position), under frequency and low flow reactor trips are blocked and will allow the reactor to remain untripped, and therefore at 10% power, on loss of offsite power.
The FSAR in reference 5, item 7.2.2.1.2d which describes this permissive provides no safety evaluation of the consequences.
Accident Analysis in Reference 7, section 15.2.9 for " Loss of Offsite Power to the Station Auxiliaries" is based on protection provided by these trips which are now blocked, and no evaluation is provided to show an acceptable RCS responsa under these particular ci~rcumstance.
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 " Comp ete 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, that below P-7, a number of other reactor trips are also blocked including Pres-suri:er Water Level-High, Pressurizer Pressure-Low and Pressuri:er Pressure-High]
The situation is one in which Condition II, III and IV occurrences are not protected in accordance with GDC 20, Protection System Functions:
"The protection 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 J a result of an:icipated 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 ce followed by a Condition III or IV occurrence in the course of protective actions.
06/01/84 6
Revision A
The licensee should evaluate the restoration of reactor trip on " low flow" trips down to and including MODE 2 (MODES 3-5 are discussed later) to be in conformance 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 Operating Occurrences and Postulated Accidents are currently satisfied, making 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 Level-High and Pressuri:er Pressure - Low are also blocked under these condi-
~
tions.
Steam Generator Water Level-Low Low remains available together with Auto-initiation of AFW pumps.
Steam Generator High Hign 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(1) Concerning Block of Pressurizer Pressure Low - Reactor Trio:
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 pressuri:er pressure low - reactor trips are used as both primary and cack l
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 i
these situations, earlier reactor trip on pressuri:er pressure low - is depended upon instead of the later reactor trip on pressuri:er pressure low - (Safety Injection).
The worst situation for most of these accidents is that of maximum power level reference 7, Table 15.1.2-2.
No evaluations are provided for :ero 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 MCOES 3-5 are discussed later). With the proposed TS reactor trips for these breaks would be delaved to be initiated later by the ESFAS (SI) related signals.
The licensee snould provide a safety evalution of these circumstances and wnics is not based upon arguments relating to probability of the events.
The evaluation snould provide 06/01/84 7
Revision A
t.
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 11 vel.
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 henc's its validity with respect to the absence of the signal caused by P7.
cil) Concerning. Block of Pressurizer Water Level-High Trip-i This pressuri er 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 tne normal control range of the reactor (without steam dumo) with the expectancy of no values exceeding normal control band (and thereby not approaching T.S. 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 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 could play a primary or backuo role in such circumstances.
For example, refer-once 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 througn 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 la hour or longer, but T.S. Table 4.3-1 shows a channel check on only a shif t basis.
Further, a single channel failure to low could cause overfill of the pressurizer (through the level control system) and with subsequent permissable failure of a second channel could remove the alarm expected fron. 2 out of 3 so that no alert is given the operator which would be contrary to the reouirement 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 06/01/94 a
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 f t 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-Hi~gh trip below P-7, including all the concerns discussed above.
This item can be interpreted as a ge'neric issue.
This'cou1d be considered non-conservative in respect to Regulatory Requirements because of the absence of 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 Con P-10 and P-13] introduces new events to evaluate for safety.
This r9 quires related Safety Analyses Limits and the Licensee shall advise what these are for each of P-10 and P-13 and how these are combined for P-7.
Item 18(f).
Propased new item:
High Power Reactor Trip on Turbine Trip; Black by absence of P-8.
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, 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.
Item:
Loss of " POWER" 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 emission from the Tecn-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.,
f P-7, j P-11 with manual blocka-9 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-OPERABILITY MCOES which are not pre-determined by Plant Logic.
06/01/84 9
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SECTIvv 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 Instructions require that boron concentration be increased to at least the cold shutdown boron concentration before cooldown is initiated.
This requirement insures a minimum of 1% delta k/k shutdown margin at an RCS temperature of 200*F " This is used as a means of protecting against NON-LOCA Accidents during startup and shutdown.
Since this' proposal to increase boron concentration is a limiting condition for ' operation required for safe operation of the facility from and including MODE 3 down to and including MODE 5, please advise why this does not appear in the Technical Specifications in accordaace with 10 CFR 50.36(c)(2).
?.S. Paga 3/4 1-1 and 2 specifying a' shutdown margin of 1.6% delta K/K over MODES 1 through 4 should be modified to exclude MODES 3 and 4, and SHUTOOWN MARGIN T should be changed from >200*F to >,557*.
4yg A new T.S. Page 3/4 1-2(a)'should be added for BORATION CONTROL SYSTEMS in MODES 3 through 5, from T
< 557'F through 140*F, providing that the boron concentration in the RCS 0911 be increased to a value which will give a shutdown margin of 1% delta K/K at 200*F.
Safety Signficance:
These actions are necessary to bring the safety status of the plant into conformance with the Licensing Basis. Without this, the plant is in a less than conservative MODE which has not been evaluated.
Further, it appears that OPERABILITY REQUIREMENTS of Table 3.3-1, REACTOR TRIP 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 3yg _ 557*F.
for T The proposed T.5. 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.
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 range high flux level alarm.
This is effectively an LCO that is not included in the proposed T.S.
The proposed T.S is non-conservative with respect to the Licensing Bases.
The Licensee shall evaluate our concerns under this Section 3/4.1.' and propose.
TS Page 3/4 1-6.
MINIMUM TEMPERATURE FOR CRITICALITY The existing minimum temperature for criticiality (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
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 zero 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, f
To be within the Licensing Basis, this TS Section 3.1.1.4 should therefore provide that the Temperature for criticality [at zero power] shall be a set i
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 Licensee shall evaluate, including the above concerns, and propose.
Section 3/4.1 2 BORATION SYSTEMS T.S. Page 3/4 1-7:
Concerning "B0 RATION SYSTEM, FLOW PATH - SHUTOOWN.
APPLICABLE MODES 5 and 6:
The current T.S. requires an (unidentified) charging pump to supply Boron to the RCS.
Current Licensing constraints on ECCS operation discussed under Section 3/4.5 Emergency core cooling systems" require that only one centrifugal 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 l
this case] the 10 CFR 50 Appendix G Limit is not reached.
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-condition for the " uncontrolled Baron Oilution Event" is that "the charging pumps are inoperative."
These circumstance permit only one charging pumo, which must be a centrifugal pump only, in operation from " standby (at 1000 psig/425*F) through to MODE 5";
therefore the term SHUTOOWN in the title and the APPLICABLE M00ES 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 ECCS and other operations under these circumstances."
The proposed T.S. is non-conservative in respect of the Licensing Basis.
The Licensee shall evaluate and propose.
l 06/01/84 11 Revision A
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T.S. Page 3/4 1-8.
Concerning:
" FLOW PATHS - OPERATING" in APPLICABLE MODES 1 2, 3 and 4.
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 shal1 evaluate and propose.
The ACTION statement should be revised to be consistent with the 8 oration Requirements adopted out of item "Section 3/4.1.1" of this report.
T.S. Page 3/4 1-9 concernino:
CHARGING PUMP-SHUTOOWN Consistent with the work of the previous TS Section 3/4 1-7 of this report, this title should 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.
APPLICA8ILITY is changed from IEE5 5 ana 6 to MODE 3 (at < 1000 psig/425'F), 4 and 5.
MODE 6 is deleted.
Surveillance Requirements 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, Cand 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 tnat the motor circuit breakers are secured in the open position by being ooened, 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 tacced 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, T.S. Page 3/4 1-10 Concerning:
CHARGING PUMPS - OPERATING AND APoLICABILITY M00E5 1. 2, 3 and 4 This is directly related to the proposed changes under Item T.S. P3ge 3/4 1-8 of this report.
Consistent with that discussion, the title should be changed to delete MODE 4, and MODE 3 conditioned to (down to 1000 psin/425'F) ftem 4.1.2.4.2 under SURVEILLANCE REQUIREMENTS does not now apply since it 300*F which are not now covered by this section, being refers to conditions y,f 1000 psig/425'F in MODE 3.
Ilmited to a minimum o The same comment applies to footnote (_, concerning one only centrifugal charging pump at 1300*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
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[
T.S. Pace 3/4 1-11 Concernino:
BORATED WATER SOURCE - SHUTOOWN This title (and related Applicability MODES 5 and 6) should be changed to 80 RATED 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 eariter since this page refers to borated water sources for situations there described.
Additionally, (by letter to reference 17] the Licensee has committed to provide j
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
)
. be included under 4.1.2.5.a.4) in which the borated water source would be demon-I strated OPERABLE by verifying minimium levels in the system.
Furtner, an additional surveillance should verify the availability of, Level, Detection-(2 indicators / tank) and related high, low and low-low level. alarms.
l Clarify whether the LCO values proposed are Safety Analysis Limits or Set Point l
Values.
An appropriate modification may need to be made to the Baron Concentrations and volumetric requirements in the Boric Acid Storage System in these MODES 3 (1000 psig/425') through 5 to provide for the increased Baron Concentrations l
required from the Licensing Basis in these MODES discussed in this report under l
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 l
wnen reference 8, page Q212-57, revision 25, under Case-3 provides that in l
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 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> from the RWST and subsequently the RHR pump and heat exenanger would re-circulate and cool from the containment sump.
Would not this require an unchanged requirement from MODES 1 through 4 of at least 372,100 gallons.
The proposed T.S is non-conservative in respect to the Licensing Basis.
The l
Licensee shall evaluate, including all our concerns above under T.S. Page 3/4 1-11, and propose.
T.S. Pace 3/4 1-12 concernino:
BORATED WATER SOURCES - OPERATING (in related Acolicaole MODES 1. 2. 3 and 4)
This title, and related applicability modes, should be changed to:
20 RATED WATER SOURCES - 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 i
describad.
Additionally, (by letter to reference 17] the Licensee did ccmmit 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 recuirements should be included under 4.1.2.6.a.4) in which the borated water source wculd be demonstrated OPERABLE by verifying minimum levels in the system.
l 06/01/84 13 Revision A
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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 LCO values given are Safety Analysis Limits or Set Point Limits.
An appropriate modification may need to be made to the Baron 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 cor.cerns under TS Pages 3/4 1-12, and propose.
T.S Page 3/4 1-13a.
P"coosed concerning:
INSTRUMENTATION IN MODES 3, 4, 5 sna 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 rsnge high 91ux 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 level or lower any time the plant is in the cold shutdown Mode.
The set point is to be checked and adjusted on a weekly basis if in the cold shutdown mode for an extunded period."
This SER requirement has not been provided in the Technical Specifications.
Please discuss provision under a proposed new item under Section 3/4.1 REACTIVITY CONTROL SYSTEMS, entitled " INSTRUMENTATION" in which these require-ments would be proposed for Applicable M00ES 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 MCOTS, it should be provided in this context also as discussed above.
The proposed T.S. is less conservative than the Licensing Basis.
The Licensee shall evaluate and propose.
T.S. Psce 3/4 1-20 Concernino:
SHUTOCWN R00 INSERTION LIMITS T.S. Page 3/4 1-21 Concerning:
CONTROL R00 INSERTION LIMITS a)
Soecifications for limiting conditions of operation on the uositions of these movable control assemblies apply only to M00ES 1 & 2.
There is no Technical specification on positions in MCOES 3-5 althougn T.S. Page 3/4 1-18 l
concerning " Position Indication system - snutdown" requires operacility of a l
Rod Position indication system in MODES 3 through 5 when the reactor trip system breakers are in the closed position.
06/01/84 14 Revision A
.o 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 consequences of 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 Baron Oilution events are specified from Refueling through to Hot stanoby as All Rods Out (Mode 6, Refueling) and, All Rods In with Most Reactive Rod Stuck Out, for Hot Standby through Cold shutdown.
- Further, app 1tcants 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 15.2415, revision 10.
Reference earlier discussion on TS. Pages 3/4 1-1, 2,and 2 a.
This is an LCO and should appear in the proposed T.S.
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 acceptable safety evaluation is submitted to show the contrary.
In general, alr1, the same concern aoplies to any other Condition II, III and IV occurrence wnich can lead to a return to power in these Modes.
Until these 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 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.
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 5HUTC0hN and CONTROL RCD positions during MODES 3 througn 5.
Without this, the prooosed T.S are non-conservative with respect to the Licensing Basis.
b)
Overpower (AT) and overtemperature (AT) protection systems incoroorste automatic Ifmits (Rod stops) on control rod insertion to maintain Jafety Analysis Limits on " Power Distribution" in the Reactor Core during power runback.
Please advise why there are no surveillance limits and requirements for these l
Rod stops in your Technical Specifications to meet the requirements of l
10 CFR 50.36. Without these, the proposed T.S. must be considered non-l conservative.
06/01/84 15 Revision A t
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Section 3/4.2 POWER O!STRIBUTION LIMITS Section 3/4.2.1 THROUGH 3/4.2.4 POWER O!STRIBUTION LIMITS RS8 has not reviewed these sections on the understanding that they are the primary responsibility of Core Performance Branch.
Section 3/4.2.5 DN8 PARAMETERS AND TABLE 3.2-1 DN8 PARAMETERS The current information does not adequately represent all those perameters necessary to ensure " acceptable" RCS operations, including DN8, under all Licensing Basis Conditions II, !!! 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 representee 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 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 cover a larger N' of variables than those presented in Table 3.2.1, in comoina-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 bqen discussed.
b)
Concerning Table 3.2-1.
The value for Reactor Coolant System T given a
as 1 593*F is not in accordance with the FSAR, reference 3, Figure S p3 1 cnere a value of 588.1*F is given as the programmed i for RATED THERMAL POWER Conditions.
Pleaseexplainthedifferenceand$$1ainwhysetpointand allowable values should not be provided.
As a Setpoint, the proposed TS value is non-conservative with respect to the Licensing Basis.
Please explain why a related power level has not been ascribed to this temperature.
Please explain why programmed T of 557.0*F (also reference 3, Figure 5.3.3 1 has not been given for Zero pow $Noperation (Reference again our Section 2.1.1 item f).
l c)
Concerning Table 3.2-1 Pressuri:er Pressure.
Please explain the basis for the given value of 3 2230 psia when information in reference 20, Table 4.1-1 (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 quotep a Setpoint value, or an allowable l
06/01/84 16 Revision A l
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value; both should be available. As a Setpoint, the proposed T.S. value is non-conservative with respect to the Licensing 8 asis for DN8R, and conservative for l
overpressure protection.
t l
d)
Why should not programmed 7,,, be provided under T.S. Section 2.1.1 e)
Why should not Pressurtze Pressurer be included both under T.S. Section 2.1-1 f
and T.S. Section 3/4.4.3 Pressurizer.
l f)
As discussed in Section 2.1.1, Subsection f, additional parameters necessary
[
to the validity of Accident Analyses in Section 15 include Pressurizer Level (See our review under Section 3.4.4.3, T.S. Page 3/4 4-9) and Steam Generator
+
i Levels under Section 3/4i4.5 T.S. Page 3/4 4-11).
j CONCLUSION
{
The ' parameters proposed by the T.S. as "0N8R PARAMETER" under TA8LE 3.2-1 are an I
incomplete set and inadequately defined in terms of Set Points. Allowable L
Values and Safety Analysis limits.
All this necessary information is available from the existing Licensing Basis and their incomplete and inadequate repre-sentation creates a non-conservative situation with respect to the Licensing Basis.
The Licensee shall evaluate and propose.
This is only partly a generic problem arising from an inadequate representation in the W STS.
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TABLE 3.3-1 REACTOR TRIP SYSTEM INSTRUMENTATION T.S. page 3/4 3-2.
Item 6c:
Source Range, Neutron Flux Does this channel provide an alarm only function, or an alarm plus trip function.
During shutdown in MODES 3, 4 and 5, with reactor trip system breakers open, 1
Source Range, Neutron Flux, channel operability requirements specify only one channel operable, and if this same channel is being used to meet the Baron dilution alarm requirements of proposed T.S. Page 3/4 1-13 (a), then it is not in accordance with the Baron 011ution Requirements of the FSAR for which at i
least 2 operable channels'would be required; reference 8, page Q212-24, item 212.58.
The Licensee shall evaluate and propose.
Currently, this' appears non-conservative.
Item 6a:
This Technical Specification concerning Operability of the Source Range Neutron Flux is unclear.
It species operability of the Source Range Neutron Flux trip Below the P 6 (intermediate Range Neutron Flux Setpoint) i during startup in MODE 2; the Licensee shall advise if this " start up" channel is required to be Operable to get Reactor trip ir MODES 3, 4 and 5.
Items 1 through 5:
The FSAR, Reference 5. Table 7.2.1-4 1 of 5 shows the i
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 from 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 l
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 I
Hign Neutron Flux Trip) because the trip function is not assumed to function in Accident Analyses.
Please note further that this position is followed through in Table 3.3-2 Items 5 and 6 in that a response time is not provided i
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 l
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 Soec proposed for it.
At this moment, absence of OPERABILITY requirements for the Power Range Neutron Flux Trip, Low Sotpoint, in MODES 3 through 5 would apoear to constitute a disparity with the Licensing Basis FSAR and in a less than conservative manner.
The Licensee shall evaluate and propose, those safety-related neutron Flux trips thich would be appropriate to use and available to trip the reactor for any of I
those events causing a return to power and under circumstance in whfen 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, the Licensee shall define and T.S. those conditions and parameters in accordance with 10 CFR 50.36, wnien would prevent any such event occurring.
t 06/01/84 18 Revision A
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 on movable control assemblies, and which instrumentation consists of only two j
channels.
Also for circumstances in which all normally available other backup trip functions such as pressurizer pressure - high and low, and water level high and " low reactor coolant flow", are not spectfied to be OPERA 8LE in 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 not 3.
Occurrences using these reactor trips include events which can be initated from suberitical Zero Power in MODE 2 (leference 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 Condition k f and Shut Down margin is required between MODES 2 & 3, how can the Licensel }ustify removal of these trips on entry into M00E 3 in whien the l
only difference in RCS conditions is a marginal reduction in temperature, from the Prograr,med No Load T,yg.
Item 11:
Pressurizer Water Level - High l
Operibility considerations from MODE 2 down to and including water solid con-ditions in the RHR MODE are discussed under Section 2.1.118 c(it.) with a proposal that exclusion of this trip for all these MODES is non conservative in rescoct to 10 CFR 50, GCC 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 items 2 21 (solected).
Items 2-21 (selected):
i Items 2, 5 and 6:
Power Range, Intermediate Range and Source Range Neutron Flux Trips i
l l
Item 9:
Pressurizer Pressure Low Item 10:
Pressuri:er Pressure High i
Item 11:
Pressuri:er Water Level High Item 12:
Low Reactor Coolant Flow Item 14:
Undervoltage Reactor Coolant Pumps Item 15:
Underfrequency Reactor Coolant Pumos Item 21:
(Proposed) loacter Coolant Pump Breaker Position Trip.
06/01/34 19 Revision A
{
At this time, in MODE 3, 4, and 5, the proposed Technical Specifications for i
the plant do not provide any neutron flux trip for Accident Analysis require-l ments, although the FSAR would require the Power-Range Neutron Flux Trip, Low Setpoint; no insertion limits on movable control assembifes, Reactor Coolant l
Pump (RCP) operability requirements permitting less than four (4) RCPs in
)
operation, a Baron 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 j
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 protections adequately l
protect the plant against any of the appropriate Condition II, !!! 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 simole solutions are possible, but at this time we do not have available an acceptacle analysis and evaluation justifying the proposed T.S. position.
l The Licensee shall provide an analysis and evaluation of the circumstances under applicable Conditions !!, I!! and IV occurrences in MODES 3 througn 5 1
1 for an appropriate se'. of Technical Specification requirements, to ensure conformance to Acceptable Regulatory Criteria and from this he will establish an appropriate range of Reactor Trio System Instrumentation to Safety Related i
Requirements.
Theevaluationshallbeundertakeninconjunctionwithour t
concerns for current Technical Specifications under Section 3/4.4.1 REACTOR COOLANT LOCPS AND COOLANT CIRCULATION of this report.
Items:
12 Low Reactor Coolant Flow Trip 14 Undervoltage - Reactor Coolant Pumps 15 Underfrecuency - Reactor Coolant Pumps 21 (Proposed) Reactor Coolant Pumo Breaker Position Trio All these Reactor Trip Functions concern potential for a loss of Reactor Coolant Flow.
The proposed T.S. deletes all operability requirements in I,
MODES 3 througn 6.
(It also deletes in MCDE 2, but this has been discussed l
esrlier under TABLE 2.21 items 18.b.a and 123 and 12b).
'ne have discussed our related concerns and requirements for analyses and evaluations in M00Ei 1, 4 and 5 under Items 2-21 (selected) above.
A loss of Coolant Flow in the RCS places the plant in an Emergency Ocersting Mode.
Please advise therefore why such an event should not automatica11v trio the Resctor in MODES 3 througn 5 with the 3eron Concentrations being considered for the proposed Technical Specifications.
'nhy should we not use the reactor trip as a device to ensure complete shutdown of all movable control rods during sny time that a minimum set of RCPs in accordance with operability requirements of tne T.S., are not availacle since RCPs may be required for accident mitiga-tion in MCDES'3 througn 5 as aporooriate.
fne Licensee shall eviluate and
- prooose, i
06/01/34 00 Revision A i
1 o.
l Item 13:
Steam Generator Water Level - Low Low:
l l
Why should not this be required for MODES 3, 4 and 5 (with closed loops) to embrace the possibility of a return to nuclear power under these conditions.
Further Steam Generator Operability is also required in these Modes to remove decay heat, and Low-Low level alarms are derived from the steam 1enerator low-l low instrument channels.
Reference 5. Figure 7.2.1-1.
The Licensee shall evaluate and propose.
Item 17:
Safety Injection Input From ESF.
l 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 5.!) is not required to be.0PERABLE in MODES 3 and 4 Why is reactor trip not required in these MODES when Taule 3.3-3' for ESFAS Instrumentation, and more particularly Func-tional Unit 1, including Reactor Trip, shows operability requirements down to and including MODE 4 Further, the licensing basis provides that S!, including reactor trio, be initiated automatically and manually down to MODE 4; see Licensing Basis information in later Section 4.5, EMERGENCY CORE COOLING l
SYSTEMS, under GENERAL, of this review.
t This proposed T.S requirement is therefore non-conservative with respect to l
the Licensing Basis which requires that Reactor Trip on ESFAS (or SI) be l
Operable in MODES 1, 2, 3 and 4.
The Licensee shall evaluate and propose.
i The Licensee shall evaluate the safety consequences of the fact that in the event of a Main Stream Line Break below tne P-11 interlock, Reactor Trip will 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 wnich will cause the reactor tript if the break is inside containment will Containment Pressuro High initiate reactor trip within an acceptable time. What are the consequences of a small to intermediate size break inside containment where, such Containment Pressure - High may not occur.
We appreciate tnat Source Range and Intermediate Range Nuclear Flux trips could trip the reactor under these circumstances, on any return to power, but their current croposed status as not being necessary for protection because they are not required in the Safety Anal-yses would leave only the Power Range Low Setootnt Trip, and related resulting power levels of 35% as a Safety Analysis Limit would be unacceptaule without a substantive analysis of the event.
Please comment in terms of Resctor Trip l
System Instrumentation Requirements to meet these circumstances.
The procosed T.S is non conservative in respect of Regulatory Requirements in meeting these circumstances; the Licensee shall evaluate and propose.
Item:
Concerning Proscribed Values For ". RATED THERMAL POWER DURING STARTUP l
PODE 2) AND POWER OPERATICH (MCDE.)
We note that oDerability requirements for Reactor Trip System CDerstion ano1 exoressed in terms of MCOES 1 and 2 are inaccurate and co not recrosent tre l
06/01/04 21 Revision A i
l
- l. -
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 Power.
~
In actual fact,the operability positions defined in Table 3.3-1 reflect an inter-face between MODE 1 and MODE 2 determined by Permissive P-7 at a nominal 10%
Rated Power Level.
Further, in this review, under Section entitled TABLE 2.2-1, REACTOR TRIP SYSTEM INSTRUMENTATION SET POINTS, item 18 c(fii) we have identified the need for Safety Analyses Limits for P-10, P-13 and in combination for P-7, so that the outer Limits of Power level of this safety control logic can be identified for safety evaluation purpose's.
For example, the Safety Analyses Limit used in the FSAR for the Power Range, Neutron Flux - Low Set Point is + 10%
on the Set Point of 25% to give 35% as the conservative outer limit.
If this same (total channel error) margin was applicable to both the P-10 and P-13 channels to give a P-7 Safety Analysis Limit of 10% + 10%, i.e, 20% RATED THERMAL POWER, then the.importance to related safety-related issues is substantively. increased.
The discrepancy identified is non-conservative and important on at least 2 counts:
1)
A non-conservative discrepancy between the fundamental maximum T.S. Limit of 5% power level in MGDE 2 as given on T.S Page 1-9, Table 1-2 and the nominal value of 10% with a real Safety analysisr Limit of 10% plus a Total Channel Error as yet unspecified.
2)
The elimination of most reactor trip Functions (and many ESFAS Functions)
~ t this non-conservative power level without a separate comprehensive a
Safety Evaluation with respect to Regulatory Requirements and the existing Licens,ing Basis.
The Licensee shall evaluate, including our. concerns enpressed above, and propose.
06/01/84 22 Revision A
TABLE 3.3-2 REACTOR TRIP INSTRUMENTATION RESPONSE TIMES 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 in MODES 3 through 5, the only remaining operable trips are those using Source range neutron Flux and they also are not required by Safety Analyses.
Under 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 safe condition in respect to Condition II, III and IV Occurrences in MODES 3 through 5.
Until this evaluation and proposal are accepted, the Licensee shall have a Safety Related Manual Trip System to assist in meeting' minimum Regulatory Requirements in 10 CFR 50, APP. A. III. Protection and Reactivity Control Systems, and the Licensee shall evaluate and propose as a priority issue..At this time, the proposed T.S is non-conservative in respect to Regulatory Requirements for 10 CFR 50, App. A. III.
Items 5 and 6:
Intermediate Range and Source Range Neutron Flux Trips.
As indicated under item Table 3.3-1, items 1-5, these items are proposed as not being protective actions necessary for the FSAR. Analyses already requested will provide a base for determining whether those trips are necessary to pro-tect the plant in MODES 3 through 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 be evaluated by the Licensee with current FSAR requirements for the Safety Analyses Limit of the power range, neutron flux, low setpoint trip wnich will be required to be OPERABLE.
The current proposals for these trips is non-conservative with respect to other proposals in the T.S; the Licensee shall evaluate and propose.
s Item 8: Overpower AT.
No respdnse time is provided by the Licensee who proposes that a T.S. on this is Not Applicable.
Please comment on the fact that this reactor trip is proposea 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 ;econds (including a transport time of 2 secs) and which is confirmed by Reference 7, Table 15.1.3-1 [alongside Overpower AT].
k The proposed T.S is non-conservative with respect to the Licensing Basis.
The i
Licensee shall evaluate and propose.
Item 9:
Pressurizer Pressure - Low 06/01/84 23 Revision A
Item 10:
Pressurizer Pressure - High The TS specifies a Response Time of <2.0 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.
Item 11:
Pressurizer Water Level - High No response time is provided because it it considered Not Applicable (NA).
The trip is shown as having a protective function for two Condition II occurrences in Reference 5, Table 7.2.14 (4 of 5) and a potential protective function in a Condition IV occurrence in Reference 7 page 15.4-13, item 16 c.
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-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 respect to the Licensing Basis.
The Licensee shall evaluate and propose.
Items 8 & 11 General
' Although the above two items are not apparently the primary reactor trips used as the basis for calculating protection in the Accident Analyses in reference 7, those Analyses represent a limited 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 l
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.
t 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 Ohis particular event at full power "A turbine trip and reactor trip are actuated when the steam generator 1..el 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 rea-tor trip is actuated.
This trips the turbine and prevents excess cocidown thereby expediting recovery from the incident.
06/01/84 24 Revision A l_
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-able).
This is incorrect as a separate Reactor Trip is an essential part of all,ESFAs functions during which safety injection is initiated.
The requiied information 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.
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 < 2 secs.
No docketed information is available on what values were used in accident analysis, and particularly for MSLB, SBLOCA and
- .0CA 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 Pressurized 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 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 the associated response time to validate proposed T.S.
values.
Item 21, Proposed (Reactor Coolant Pump Breaker Position Trip)
As discussed earlier, under table 2.21, Item 14, this trip is provided as an adjunct to Undervoltage - Reactor Coolant Pump Trip.
The Licensee shall evaluate and propose.
06/01/84 25 Revision A
(
e e
TABLE 3.3-3 ENGINEERED SAFETY FEATURES ACTUATION SYSTEM (ESFAS) INSTRUMENTATION Item 1:
Safety Injection, Reactor Trip, Feedwater Isolation, Component Cooling Water, Start Diesel Generators, and Nuclear Service Water.
This description of Item 1 lists the various functions initiated by given signals (which are generally those initiating SI).
However, Reference 5, Figure 7.2.1-1 (8 of 16) revision 34 and Figure 7.2.1-1 (13 of 16) revision 34, shows that the term "Feedwater Isolation" used in this Item 1 is actually comprised of four (4) separate Logic Functions, namely
" Turbine ' rip", " Trip of Feedwater Pumps", "Close All Feedwater Isolation Valves" and "Close the Feedwater Main and Bypass Modulating Valves.
The term Feecwater Isolation is therefore an inaccurate term to use.
It shou,ld be removed frcm 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 Generator Water Level High-High (P14) reference 5, figure 7.2.1-1 (13 of 16) revision 34. and figure 7 of 16; revision 41.
Further, the function to be initiated by Steam Generator Water Level - High High is function 5 of the same Table which is again incompletely described and should be changed (see item 5 later) to clearly identify these same 4 elements.
Under these. circumstances, the current description for Item 1 snould delete the term "Feedwater Isolation" and Item 5 (see later) should be expanded to include an additional Functional Unit identified as Safety Injection.
Additionally, the Function " Annulus Ventilation" needs to be added to the descriptor (reference 5, figure 7.2.1-1 (8 of 16) revision 34).
Also, the function unit description " Nuclear Service Water" should include
[ isolation and startup] of Nuclear Service Water.
Item la):
Manual Initiation This should read as:
Manual Safety Injection Actuation.
[There is not a separate Manual Actuation for each of the functional units listed.]
Item ic:
Containment Pressure - High/ Applicable 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 - Law, provision is.
made that if a MSLB occurs inside containment [so that MSIV Isolation on 06/01/84 25 Revision A
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 3I is already provided in the T.S. in MODES 3 and 4.
This 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 ld:
Pressurizer Pressure-Low
' his is the same title as used for Reactor Trip on Pressurizer Pressure-Low.
T 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,yg Interlock) setpoint.
Reference 5, Table 7.3.1-3 (1 of 2) and (2 of 2) item P-1, shows the appropriate interlock for this purpose is P-11.
Item P-12 of the same Table makes no provision for this proposed T.S. position.
However, reference 5 figure (6 of 16) does not use the same manual block (at P-11) for 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 otner 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 tne correct cescription.
Item 2c:
Containment Pressure-High-High.
Operability is not required in MCuE 4.
This should be required to be consistent with the evaluation under Item 3.b.3. belcw.
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 advise why this is not necessary for safety when the previous item No.l.e.
06/01/84 27 Revision A
t.
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 7 2.1-1 (8 of 16), revision 34.
The Licensing Basis for this requirement lies 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 propo'se.
Item 4, Steam Line Isolation 4b: Automatic Actuation Logic and Actuation R'elays
~
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 requirements of Steam Line Isolation on Negative Steam Line Pressure Rate -
High, already specified in item 4d for MODE 4, are to be met.
The proposed.T.S.
is non-conservative with respect to the Licensing Basis; the Licensee shall evaluate and propose.
Item 4a:
Manual Initiation (of steam line isolation]-
1)
System 2)
Individual Operability. requirements for manual initiation of Steam Line Isolation are not
~
required by the current T.S. in MODE 4.
This however will be necessary to allow the operator to manually isolate small breaks which do not activate the Negative Steam Line Pressure Rate - High signal or the Containment Pressure-High High signal.
The proposed T.S. is non-conservative with respect to the Licensing Basis; the Licensee shall evaluate and propose.
Item 4d:
Negative Steam Line Pressure Rate - High Operability requirements are given-as MODE 3 and 4.
MODE 3 should be con-ditioned as MODE 3# indicating it is only available below P-11 Interlock.
The Licensee shall evaluate and propose.
06/01/84 28 Revision A
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]/ APPLICABLE 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 Valves 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 availablity of Main Feedwater Pumps during these MODES.
Plant Operating Procedures which permit Main Feedwater Pumps to be available can cause An Excessive Heat Removal 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 Logic.of reference 5, figure 7.2.1-1, (13 of 16), revision 34, involving signal inputs:
Steam Generator Hi-Hi P-14, Safety Injection, Reactor Trip P4, and Low T,yg would need to be carefully 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 T3 g (during cooldown) to effect Feedwater Isolation, and Closure of Modulating Valves, as an inbuilt protection against such circumstances.
The proposea T.S. does represent a non-conservative position in respect to the t
!~
Licensing Basis, as there is no prerequisite that Main Feecwater 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 Injectic' [to effect Turbine Trip, Feedwater Pumo Trip, Closure of Feedwater Isolation Valves and Closure l
of Feedwater Modulating Valves]/ Applicable MCDES PROPOSED
}
AS 1, 2,'3 and 4 This trip is relocated from Functional Unit 1 to Functional Unit 5 in accordance with our earlier reviews under Item 1C and Item 5.
l
[
06/01/84 29 Revision A i
1 l
?*
OPERABILITY is required in all Modes 1, 2, 3, 4, because SI protection has been found necessary within the Licensing Basis.
The protection was already intended in the proposed T:5. 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 pa'ge 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 2
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 car. only be achieved by having available the Automatic Actuation Logic and Actuation Relays, and all related ESF equipment so that the automatic logic can be initiated manually.
The licensee shall evaluate and propose.
The proposed T.S. items '7'a 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 for operable RCS systems in MODE 4.
The proposed T.S. items 7a througn 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, including consideration of our related review.
Item 7.a:
AFW/ manual initiation Item b:
AFW/ Auto Actuation Logic and Actuation Relays:
Operability is currently not required in MODES 4 and 5.
Operability should be provided for both modes to meet the licensing requirements, i.e., manual initiation of Automatic Actuation Logic and Actuation Relays:
reference General above.
Item 7.c.1:
Start Motor Driven Pumps:
Should be operable in both MODES 4 and 5 and especially to counter non-availability of Turbine Driven Pumps early into MODE 4 during the cooldown.
1 06/01/84 30 Revision A
1 Item 7.c.2):
Start Turbine Driven Pumps:
Should be operable in 4.
Although not capable of operating at lower tem-peratures of MODE 4, and MODE 5, it should nevertheless be available for use to counter consequences described in " General" above, including a station blackout.
Item 7.d): Auxiliary Feedwater Suction Pressure Low:
This proposed T.S description of a. functional unit is invalid.
The Functional Unit to be provided is:
d) Automatic Re-alignment of Suction Supply [This is the functional unit],on Low Auxiliary Feedwa'ter Suction Pressure [This.is the parameter caus-ing the change]
Operability requirements should identify how many AFW pumps are required to be " tripped" deficient in suction, to effect re-alignment.
The licensee should identify those instrument / control channels, and partic-ular engineering alignments, which result in a re-alignment of 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.
Item 7.e:
Start Motor-0 riven Pumps (by Safety Injection)
Applicable Modes have not been identified.
NRC proposes MODES 1, 2, 3 and 4 and 5 to meet the requirements of Item 7:
General, discussed earlier.
Item 7.e:
Start Turbine-Oriven Pumps (by SI)
This functional unit proposes that the Turbine Driven AFW pumos are started by the SI signal.
This conflicts with reference 5, Fig. 7.2.1-1 (15 of
Also, in a like manner, with related sec-tion 7.4.1.1.1.1. and reference 22, section 10.4.7.2.2.6.
Also see refer-ence 14 Section II.E.1.2 page 22-41.
It is now noted that the recent T.S. has been corrected to show that the Turbine Driven AFW pumo does not start on Safety Injection.] The Licensee shall clarify.
06/01/84 31 Revision A
Item 7.f; Station Blackout - Start Motor Driven and Turbine Driven Pumps:
Provision for operability is only in applicable MODES 1, 2 and 3.
Con-sistent with previous considerations, operability should be required in MODE 4, with provision for immediate operability from MODE 5.
~
Item 7.g: Trip of Main Feedwater Pumps (MFWP) - Starts Motor Driven Pumps The T.S. proposed only 1 channel per pump to trip.
[This is different to the FSAR, reference 22, page 10.4-14, rev. 7, item 30 which specifies that loss of all main feedwater pumps is required.
The licensee should evaluate and propose.
Applicable modes:
The current T.S. proposes Modes 1 and 2#.
Condition 2#
is an invalid MODE since # identifies the P-ll 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
~
propose.
Please advise why this limitation at MODE 2 [or 3]# is proposed and how it may relate to plant operating procedures in MODES 3 and 4 and whether this block is in conformance with regulatory requirements.
Item 8:
Automatic Switchover to Recirculation on RWST Level:
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 Sec-tion 3/4.5 ECCS under GENERAL, the licensee should evaluate the reasons for, and the consequences of, not proposing this OPERABLE IN MODE 4, and not being available in MODE 5, to counter the consequences of potential LOCAs and loss of RHR cooling in these MODES.
The proposed T.S. is non-conservative with resoect to the Licensing Basis; the Licensee shall evaluate and propose.
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 Engineerec 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.
Absence of this information makes the proposed T.S. non-conservative.
~
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:
wnat pumps etc., and if so operability requirements should be extended to MODE 5.
Lack of any clarity makes this proposed T.S. non-conservative.
The Licensee shall clarify, evaluate and propose.
06/01/84 32 Revision A a
4 I
Item 10.,ph.:
Pressurizer Pressure P-ll:
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 Pumos.
The basis provided on proposed T.S. Page B 3/4 3-2 states that:
"P-11 (i.e., on system pressure increasing to P-11 valve) --- Defeats the manual block of the motor driven AFW pumps on trip.of the main feed-water pumps and Low-Low Steam Generator level."-
The following information provides the current Licensing Basis on the particular proposed interlock P-ll in respect of AFW Pumps:
The Table 3.3-3, Item 7.c.1, in reference 5, for start of motor driven AFW
_ pumps, does not provide for the above condition.
The P-ll 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.
P 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-ll 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 g
l 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 I
absence of auto-initiation of motor-driven AFWS as now proposed.
If the Licensee wishes to pursue **11s 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 protections available and the time frames which must allow for all other actions, e.g.,
-isolation of a ruptured SG is exoected 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 l
E-
We note that in his submittals concerning this matter, dated March 9, 1981 concerning TMI items, the Licensee states that "the turbine driven auxiliary feedwater pumps do not have a bypass feature." Yet we also note on his T.S.
page 3/4 7-4 that the Turbine Driven pump is not required to be operable when steam generator pressures are less than 900 psig; this would require only approx. 20 mins. into standby cooldown to achieve.
The result is that there would be absolutely no automatic supply of feedwater for an_y event beyond approx. 20. min into cooldown.
At this time, the current Accident Analyses in the Licensing Basis FSAR support the necessity for not using the current bypass for the Motor-0 riven Pumps.
The Licensee shall advise wha.t safety-related reasons require that he must bypass automatic startup of the motor-driven auxiliary feedwater pumps on top,of both main feed pumps, and on SG Low Low-Level in the final. stages of
.i plant shutdown.
Also, what prevents him from installing automatic restoration on receipt of the related protection signal.
Item 10.b; Interlock; Low-Low Tavg P-12:
Applicablo MODES are 1, 2, 3.
Reference Item Table 3.3-4, Item 10b, of this document.
Since Interlock P-12 effectively provides ~and limits steam dump capability, including accidental blowdown, by constraining it to 3 cool down dumps to the condenser; why remove this interlock in MODE 4 and 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 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 propose.
Until Safety Related Isolation of Main Feedwater 06/01/84 34 Revision A
)
Containment Isolation Valves is included in the T.S., this proposed T.S.
must be considered non-conservative with respect to Regulatory Requirements.
Item 11 proposed:
There is a need to add a new Functional Unit not addressed in the current T.S., but which is a part of ESFAS.
This is:
"Close All.:eedwater Isolation Valves" and "Close the Feedwater Main and Bypass Modulating Valves" See reference 5, Figure 7.2.1-1 (13 of 16) revision 34 for the related.
unique' control logic.
This Function is initiated by:
lla.
Reactor Trip P-4, and Low Tavg.
11b.
Reactor Trip P-4, and Steam Generator Level - High High P-14.
11c.
Steam Generator Level - High High P-14 (see 5 abcve) 11d.
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.
Operability for 11c and 11d would be by reference to items 5, Sabc.
TABLE 3.3-3:
TABLE NOTATION The uncertainty of the notation under ## is discussed in Item le earlier.
Please amend as required in accordance with the related resolution.
06/01/84 35 Revision A O
t.
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-scribed as being applicable to McGuire Unit 1, 50-369.
At this time, the 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.
ItemNo.f:
2 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 1&
Modify the description in accordance with our earlier comment under Table 3.3-3 Id to:
Pressurizer Pressure - Low (Safety Injection)
Item 3c.4 (Proposed):
Reference 5, Figure 7.2.1-1 (8 of 16) revision 34 shows that " Containment Radioactivity" initiates containment ventilation (Purge and Exhaust) isolation.
Please explain why it is not included as, e.g., a proposed Item 4).
The pro-posed T.S. is non-conservative with respect to the Licensing Basis.
The Licensee 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 let 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 conservatisms, it is clear to delete the - in -100 as the description reads :
" Negative Steam Line Pressure Rate - High so that T.S. values should read as 100 osi and 110 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-
Please discuss the logic of the values in reference 18.
A Trip Set Point of a negative rate of 110 psi with an allowable value of 100 psi (both with a time constant of 50 psi) would provide that an earlier isolation of the MSIVs is less conservative, and this is not so for the MSLB event.
The expectations are that negative rate for the allowable value would be higher than for the Set Point.
Please clarify.
Further, the same reference 18 Table 3-4, column 12, states under 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 i
Break below the P-11 block point (instead of by Steam Line Pressure - Low) 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.S. 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-conserva'tive.
The Licensee shall evaluate and propose.
Item 5:
The description of this Functional Unit should be revised and clarified to our recommendations under Table 3.3-3, Item 5.
Item Sc:
Proposed new item as " Safety Injection" This should be included in accordance with the evaluation under 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.
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 ano Turoine 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
a 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, Section 10.4.7.2.2 states that " Travel stcps 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 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 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 gpni to the intact steam e
generators one minute after reactor trip.
At 30 minutes following the rupture, the operator is assumed to isolate the auxiliary feedline to the ruptured steam generator which results in an increase in
~
injected flow of 80 gpm."
The sequence of events in.the accident evaluation in Reference (7),
Table 15.4-1 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 20 secs. later, and auxiliary feedwater [at 450 gpm] is delivered to tne intact steam generators in 61 sec.
It appears, based on the above information, that on SG low-law 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 oesigned so that under normal operation, collapse of SG level on Turoine 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 i
other concerns on this event noted later in Sections of tnis review.
06/01/84 38 Revision A 1
i
Item 8:
Automatic Switchover to Recirculation The Licensee shall provide the basis for the set point values of the RWST levels specified. What are the allowable values for [ drift and] total channel errors and the related Safety Analysis Limit.
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 OFFSITE POWER TO THE STATION AUXILIARIES describes a set of Reactor Protection System and Engineered Safeguards Features Actuation Respo'nses for the Plant,' to ensure,its safety.
Why i.s.this particular set of ESFA's Functional Units and related Instrumentati6n Set Points not provided in this item under Table 3.3-4?
Absence of this information makes the proposed T.S. non-conservative.
The Licensee shall evaluate and propose.
Item 10a: 'ESFAS Interlock Pressurizer Pressure, P-11.
Actuation of this interlock substantively reduces ECCS protection against Conditions II, III, and IV Accidental Occurrences.
The FSAR has analyzed the consequences of this reduced level of protection for a limited number of these occurrences and this has been 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 <1955 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,yg-P 12 The basis for this interlock on T.S. Page B 3/4 3-2 states that:
i "On decreasing reactor coolant loop temperature, P-12 automatically removes the arming signal from the steam dump system." This is not substantively consistent with Reference 5, Figure 7.2.1-1 which shows that'it is the arming signal for the condenser dump valves and atmospheric dump valves..hich is removed and then with the excection 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
I
?.
A set point of 553-551*F is prov.ided.
Provide the basis for this which should be consistent with our query under earlier Sec-tion 3/4.1.1.
Boration Control concerning T.S. page 3/4 1-6,
" Minimum Temperature For Criticality."
Item 10e. (Proposed).
To 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 Appl.icable) trip.setpoint & Allowable values." However, should not this item read as:
10c. P-4-with Trip Setpoint and Allowable values defined as in Reactor Trip to Table 2.2-1, with the 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 Reactor Trip Signals ini-tiating P-4 include all items in Table 2.2-1 with the exception of " Power Range, Neutron Flux, High Negative Rate." Th'e licensee shall evaluate and propose Item 11 Proposed:
There is a need to add a new Functional Unit not addressed in the current T.S., but which is a part of ESFAS.
This is:
"Close Feedwater Isolation Valves & Close Feedwater Main & Bypass Modulating Valves."
(See Reference 5, Figure 7.2.1-1 (13 of 16)
Revision 34.)
This Functional Unit is initiated by:
a.
Reactor Trip P-4, & Low Tavg
- b.
Reactor Trip P-4, & Steam Generator Level - Hign Hign P-14 c.
Steam Generator Level - High High P-14 (see 5 above),
d.
Safety Injt"-tion (see 5 above). "
Trip Set Points would be in accordance with the related values in earlier Items 10 and 5 of this section.
06/01/84 40 Revision A
i l-Reference Item lib above, involving Reactor Trip P-4 & Steam Generator High High Level P-14.
The NRC has observed potential situations of concern involving this interlock.
NRC Safety Concern A: A review of the logic of this interlock, Reference 7, t
Figure 7.2.1-1, (13 of 16), Revision 42 shows that if a SG-Hi Hi occurs, i.
Turbine Trip, Trip of MFW Pumps, closure of MFW isolation and control valves occur, but the reactor is not tripped if the Nuclear Power Leve.1 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 6:
The existing FSAR, Reference 7, Section 15.2.10.1, Revision 15, shows that a feedwater malfunction at. full power.is not
- i l
terminated by a neutron Flux Power trip, but by a SG-Hi Hi (i.e., P-14) i 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-l 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 event of this sequence by reference to the, event of
" Uncontrolled Rod Cluster Control Assembly Bank Withdrawal From A Sub-l critical Condition," the FSAR provides no evaluation of the subsequent event including the ONBRs resulting from any restoration of reactivity before SG-Hi Hi ultimately effectively closes MFW isolation valves.
This 4
latter event from zero power can also occur at any intermediate power i
level, with and without automatic rod control, and there is currently no l
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 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 6
up to 58% power level.
[
In respect of NRC Safety Concerns A, B, and C above, we consider the pro-4 posed T.S. in respect of the related permissives and interlocks 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 4
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 T
06/01/84 41 Revision A h
7
[.
..-m-
,__._.._...,.,,.,,_m_-.,,,m m
e Steam Generator Hi Hi., the licensee should review the analyses for all other Condition II 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.
Reference Item 11(a) above, involving Reactor Trip P-4 and Low T,y Reactor Trip P-4 together with Low-T,yg causes closure of the MFW isolation valves and MFW Modulating (Control valves) thereby isolating the reactor from any faulted [on non faulted] feedwater system.
The safety significance of the parameter, Low T
, as expressed in the FSAR derives (a) from its inclusion in the ESFAS under Reference 5, Figure 7.2.1-1, (13 of ~16), Revision 34 and (b) a description in Reference 5, Section 7.7.1.7 under the title Steam Generator Water Leve'l 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."
This P-4/ Low T combination does perform a safety function in preventing 3yg excessive cooldown after the reactor is tripped, but has never been incorporated, or discussed in the Section 15 FSAR analyses (Reference 7) for this purpose.
Within the FSAR under Reference 7, Section 15.2.10.1 " Excessive HEAT REMOVAL DUE TO FEEDWATER SYSTEM MALFUNCTIONS" state that:
"An accidental full opening of one feedwater control valve with the reactor at zero power and the above mentioned assumptions, the maximum reactivity insertion rate is less than the max'imum 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 d.scharge valves on steam generator high-level."
This event from zero and higher power levels (already discussed under earlier Item 11b) is initially protected by the high neutron fluxtrip; however whilst this provides immediate protection, the main feedwater is 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
Generator Hi Hi Level to trip the MFW pumps, and together with existing Reactor Trip to provide Main Feedwater Isolation.
Or, is it necessary to depend on an earlier " Isolation of Main Feedwater" from the combination of the existing reactor trip P-4 signal already provided and a related Low T,yg.
Inclusion of the P-4 and Low T,yg interlock into the T.S. would provide more reliability in protection for this event in conformance with the diversity criteria of 10 CFR 50 Appendix A, GDC Criterion 22 in support GDC 20.
Without this, there is no diversity for protection from this continuing event.
The proposed T.S. should require T,yg Low to be incor-
. parated 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 T,yg, without T.S. considerations of Set Point, Maximum Errors, Channel Reliability, Applicability MODES and Action Statements raises concerns about the consequences of a single failure.
For example, a failure low, remaining undetected, could combine with a Reactor Trip from full power to close Main Feedwater [ containment] Isola-tion valves and Main Feedwater Modulatirig valves and cause a more severe transient than would otherwise be necessary.
The Licensee should evaluatie tne consequences of single failure 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 Ouring Power Operation,"
states that:
Spurious ECCS operation at power could be caused by operator error or a false electrical actuating signal.
Spurious actuation may be assumed to be caused by any of the following:
1.
High Containment pressure 2.
Low pressurizer pressure 3.
High steam line differential pressure 4.
High steam line flow with either low average coolant temperature or low steam line pressure.
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.
06/01/84 43 Revision A
', =
Item":
Reference 5, Figure 7.2.1-1 (2 of 16) Reactor Trip Signals The reference to Safety Injection Signal (Sheet 8) is inaccurate.
This signal is from the ESFAS and not directly from the SI signal.
06/01/84 44 Revision A
TABLE 3.3-5 ENGINEERED SAFETY FEATURES RESPONSE TIMES Item 2a:
Initiation of Safety Injection by:
Containment Pressure-High.
A value of 1 27 secs (without offsite power) is given.
Reference 5, page 7.3-8 shows that initiation time of ESFAS from this source is a maximum of I 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-11].
Other events dependent upon a SI generating signal, particularly circumstances descibed under items 3a and 4a below, shows safety analyses limits of 1 12 secs.
(with offsite power) and 1.22 secs (without off site power).
1 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 1 2 secs and this different from the FSAR, Reference 5, page 7.3-8 which gives a maximum time of 1 sec.
This value is less conservative than the FSAR and the licensee shall evaluate and propose accordingly.
Item 2c:
"Feedwater Isolation" from Containment Pressure-High The response time is given as 1 9 secs.
Reference 5, page 7.3-8 shows that initiation of ESFAS from this source is a maximum of 1 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, fr'om Containment Pressure-High, of 1 11 secs seems apprcoriate to available equipment.
There would then be a conflict between the response time of 1 9 secs in the proposed T.S. and the potential ualue of up to 11 sec from other licensing basis information.
No event in Reference 7, Section 15.1 through 4. uses this particular isolation in time Analyses.
However, this is a important factor for containment integrity during a Main Steam Line Break in containment.
The value used as the Safety Analysis Limit shall be provided oy the licensee, 06/01/84 45 Revision A
compared with proposed T.S. Item 2c and any differences evaluated, and T.S. oroposed as appropriate.
Item 2d:
Containment Iwlation - Phase A, from Containment Pressure-High The proposed T.S. values are 18(3) (with offsite power) and 28(4) without offsite power.
Reference 5, page 7.3-8 shows that initiation of ESFAS from this source is 1 sec.
Table 3.6-2 shows Maximum Isolation Times of up to 15 secs for Reactor Coolant Pressure Boundary Isolation valves.
A minimum total time to containment and isolation (for the RCPB] of 16 secs seems feasible, plus 10 secs giving 26 secs total without offsite power.
' The proposed T.S. values should be checked against those used as Sa'fety Analysis limits 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.
Item 2e:
Containment Purge and Exhaust Isolation, from Containment Pressure-High This is given as N.A.
This is not so; response. times have be used to minimize offsite consequences of any Condition occurring whilst contaih-ment purge & exhaust is being used.
This proposed T.S. is less conserva-tive than the ifcensing basis.
The licensee shall evaluate & propose.
8 tem 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 2g:
Initiation of Nuclear Service Water (NSW) from Containment i
Pressure-High This response time is given as < 65(3)/76(4 l
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 Servica Water Pumps as described in Functional Unit 1 of Tables 3.3-3 and 3.4-4 The same comment applies to superscript 4 which applies to the circum-stances without offsite power.
The licensee should propose an accurate description of these circumstances; the current description does not meet the intent.
06/01/84 46 Revision A
Reference 5, page 7.3-8 shows that initiation of ESFAS from this source is 1 sec.
No other information is available on Safety Analysis Limits because, contrary to Regulatory Requirements, this value has not been used in the l
l Safety Analysis of the FSAR in respect of AFW supplies.
In other sec-j 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, l
compare with the proposed T.S., and repropose as appropriate.
Any Occur-rences required to utilize Nuclear Service Water must be considered non-
,l conservative with respect to these values currently presented in the FSAR to Reference 7, Section 15, i
Item 2h:
Initiation of Component Cooling Water from Containment Pressure-High This respons'e time is given as 65(3)(3)/76(4)(2) secs.
l s
The description of superscript 2 under Table Notation on T.S. Page 3/4 3-33 i
is incomplete.
The licensee shall propose an accurate description of these i
circumstances including its dependence on Nuclear Service Water; the licensee should confirm that this cooling water supply information is for l
this safety related servi'ce.
~
Reference 5, page 73-8 shows the initiation of ESFAS from this source is 1 sec.
No other information is available on Safety Analysis Limits used.in the FSAR.
The licensee shall provide this information for related Condi-l tions II, III, and IV Occurrences for both on-site and offsite po'wer.
This l
information shall e evaluated and the licensee shall propose..At this i
time, considering the non-conservative circumstance with NSW AP4 supply, l
it must be presumed that any Occurrence required to utilize the Nuclear j
Service Water must be considered non-conservative with respect to the
]
values currently presented in the FSAR, Reference 7, Section 15.
Item 21:
" Start Diesel Generators" from Containment Pressure-High j
A response time of 511 secs is given.
Reference 5, page 7.3-8 shows that initiation of ESFAS from the source i
is a maximum of 1 sec.
}
No evaluation in Reference 7, uses this sensor as the prime initiator l
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, :.ams 3a & 4a belcw, show safety analysis limits of 1 10 secs for this value.
l 1
In respect of current safety anal'yses limits, therefore, it appears that l
the proposed value is less conservative than the Safety Analysis Limits.
7 The licensee shall evaluate and propose.
~
i I
,I 06/01/84 47 Revision A
=
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 Pressure-High.
Since both or either of these initiators can be available in Occurrences involving SI, and initiation times are the same, our comments and conclusion's under earlier Item 2 can be directly
- referenced for items under Item 3 in cases where the proposed response-time is the same for a given ESFAS function.
Item 3(a):
" Safety Injection (ECCS)" on Pressuri:er Pressure-Low [SI]
1 27(1)/12(3) secs are proposed.
Values of Reference 5, page 7.3-8, shows a maximum initiating time 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 reference 7, page 15.4-10, Sectica 7 where "In 12 seconds, the valves are assumed to be in their final position and pumps are assumed to be at full speed." For.the other case with Loss of Offsite Power (LOOP) "an, additional 10 secs delay is assumed to start the diesels and to load the necessary equipment onto them."
Further, this particular analysis appears to initiate the event on Pressure Pressure-Low (SI).
The proposed value of i 12 secs appears 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 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.
Item 3b:
" Reactor Trip (from SI)" on Pressurizer Pressure Low [SI]
The descriptor (from SI) is incorrect and should be deleted.
A value of 12 secs is proposed.
The FSAR in Reference 5, page 7.3-8 quotes a value of 1 i secs.
The proposed T.S. value appears less conservative tnan the Safety Analysis Limit and the licensee should evaluate and propose.
06/01/84 48 Revision A
Item 3c:
"Feedwater Isolation" From Pressurizer Pressure-Low (SI)
The proposed T.S. is 1 9 secs.
Reference our comments and requirements under 2.c. above.
t Item 3d:
" Containment Isolation - Phase A" from Pressurizer Pressure-Low (SI)
The proposed T.S. is 1 18(3)/28(4) secs.
Reference our comments and requirements under 2.d. above.
It m 3e:
" Containment Purge & Exhaust Isolation" From Pressurizer e
Pressure-Low (SI)
The,.soposed T.S. is NA.
Reference our comments and requirements under 2.e. above.
Item 3f:
" Auxiliary Feedwater" Initiation by Pressurizer Pressure-iaw (SI)
The licensee proposes NA (not' applicable).
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.1 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-Law SI The T.S. value is given as 76(1)/65(3) secs.
Our comments on 65(3) are as for our earlier 2g.
With respect to superscript (1) on 76; wny is this different to Containment Pressure High which is 76(3) when the concomitant SI signal generates the same equipment requirements.
Superscript (1) now provides for SI and RHR pumps whereas (3) did not.
Also, superscript (1), if it is to be used should include Isolation and Sta"t 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 suoerscripts (1)
( d. (3 ) an'd (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
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 1 76b)/65(2)(3) secs.
See our comments and requirements under 2h. and 3.
General above.
Item 3i:
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.licens~ee shal'1 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 I sec.
The licensee shall evaluate and propose.
Item 4a:
" Safety Injection (ECCS)" Initiation on Steam Line Pressure-Low 1 12(0)/22(4) agree with the Safety Analysis Limits These values of of the Licensing Basis FSAR.
Item 4b:
" Reactor Trip (From SI)" from Steam Line Pressure-Low.
t-Thedescriptfog(fromSI)isincorrectandshouldbedeleted.
This value of $ 2 secs agrees with Reference 5, page 7.3-8.
(
Item 4c:
"Feedwater Isolation" from Steam Line Pressure-Low The proposed T.S. is 1 9 secs.
Reference our comment and requirements under 2c. above modified by the
-fact that there appears to be a larger conflict between the response time of 1 9 secs and the potential value of up to 11 + 1 = 12 seconds from Licencing Basis Information.
06/01/84 50 Revision A
0 Item 4d:
" Containment Isolation - Phase A" on Steam Line Pressure-Low The proposed T.S. is 5 18(3)/28(4) secs.
Reference our comments and requirements under 2d. above, modified in that proposed T.S. times appear feasible with the additional delay of 1 sec.
Item 4e:
" Containment Purge and Exhaust Isolation" on Steam Line Pressure-Low The proposed T.S. is NA.
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.
Reference our comments and requirements under 3f.*above.
Item 4g:
" Nuclear Service Water" initiated on Steam Line Pressure-Low The proposed T.S. is 5 65(3)/76(4) secs.
Reference our comments, requirements, and remarks under 2g., 3g., and 3 General above.
Item 4h:
Steam Line Isolation on Steam Line Pressure-Low.
The proposed TS value is 5 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-law, 2 secs.
Further, Refer-ence 7, page 15.4-6 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 Steamline Isolation 3
Valves Closed." [In the case of Feedwater System Pipe Rupture]
The proposed TS value of 1 9 secs is therefore greater than the Safety Analysis Limit.
The proposed TS must therefore be considered less conservative for this event.
The licensee shall ev.luate and propose.
Item 4i:
" Component Cooling Water" Initiation by Steam Line Pressure-Low Proposed T.S. value is 65(2)(3)/76(2)(4)
Reference our earlier comments and requirements under 25 and 3h. above.
06/01/84 51 Revision A
^ ~ '
Item 4j:
" Start Diesel Generators" by Steam Line Pressure-Low.
j Proposed T.S. value is 5 11 secs.
Reference our comments and requirements under 21 above.
Item Sa:
" Containment Spray" - Initiated on Containment Pressure-High-Hig'i Licensee shall provide the Safety Analysis Limit and compare with the proposed value of 1 45 secs.
Evaluate and propose as necessary.
Stem 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 valves, Maximum Isolation. Time (secs),-applies only to closure from receipt of s'ignal, 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:
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.
Etem Sc:
Steam Line Isolation on Containment Pressure High-High The proposed T.S. value is 1 9 secs.
Reference 5, page 3.7-8 shows containment pressure initiating ESFAS signals I
with a 11 response time.
Item 4h. above shows fast acting stop valves I
closing in 5 secs. giving a total time of 16 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.
A comparison also with values used in assessing environmental releases frem 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 proposed T.S. value and propose as appropriate.
Until such time, the 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.
Reference the licensee to our ccmments under Table 3.3-2, Item 16 where it is shown that it is used within the Licensing Basis.
The proposed position is non-conservative with respect to the Licensing Basis.
The licensee shall evaluate and propose in accordance with our review under Table 3.3-2, Item 16.
Item 6b:
"Feedwater' Isolation" Initiated by Steam Generator Water Level-High High The proposed T.S. is 5 13 secs.
Reference 7, Table 15.1.3-1 shows that "High Steam Generator level trip of the feedwater pumps and closure of feedwater system valves, and turbine trip" is based on an ESFAS time delay of 2.0 seconds.
Table 3.6.2 of the T.S. provides isolation times of < 5 secs for main feedwater containment isolation and < 10 secs for main 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 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 approx. 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 restoration and/or overfill of the S-G to cause water ingress into the main steam lines. Additionaliy, it can have consequences of potentially 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.
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
1 e
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 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-Driven Auxiliary Feedwater Pumps" initiated by SG Level-Low Low Proposed T.S. response times are given as < 60 secs.
The FSAR Safety Analysis Limit is 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 ste'am generators is 61 secs.
The current i
. proposed T.S. value is therefore conservative'with respect to;the current j
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, t
the auxiliary condensate storage tank, or the condenser hotwell.
Each of these sources are provided with motor operated valves with control room operation.
The assured AFS pump suction is from the Nuclear Service Water System.
The A motor drive is aligned to the A NSWS header and the B motor driven pump is aligned to the B 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
~
l time for Nuclear Service Water Supply is 65 secs, assuming offsite power available and 76 secs assuming loss of offsite power whereas the Safety Analysis Limit used in the FSAR is only 61 secs.
On this basis, all l
Conditions II, III, and IV occurrences involvi.ng 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 i
Systems" that the related loading sequences for pumping equipment, alone, i'
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, 4 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
,e Item 11, below.
If this is confirmed at from 65 to 70 secs, or any longer time than used as the existing Safety Analysis Limit in the FSAR, then acceptable re-evaluation of all Conditions II, III, and IV occurrences involving AFW supply, are required by 10 CFR 50.36.
Our current evaluation is that the response times in the proposed T.S.
are non-conservative in respect of Regulatory requirements.
Item 8:
" Steam Line Isolation" on Negative Steam Line Pressure Rate-High Proposed T.S. value is 1 9 sec.
Reference 5, page 7.3-8 states that the maximum allowable time for generating the ESFAS MSIV isolation signal from a Steam Line Pressure Rate circumstance is 2 secs, the same as for item 4h. above.
Our comments and requirements therefore are the same as under item 4h.
We appreciate that this signal is generated at below P-ll, but with the existing proposed Boration Control T.S. we must continue to evaluate this value as non-conservative.
The proposed T.S. value is greater than the Safety Analysis Limit of seven (7) secs and must be considered less conservative for this event:
The licensee must evaluate this difference and propose.
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]
Proposed T.S. value is 13 secs.
Note auf comments under 7a. and 7b. above.
Although this response time may be in accordance with current plant engineering, it is not in accorcance with the existing Safety Analysis Limit for Auxiliary Feedwater Supply 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 re'sidual 15 secs supply at this time.
At this time, until the evaluation of 7a. and 7b. above is completed, we nust evaluate this delay as non-conservative with respect to currently lused 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 Rec.irculation" on Low RWST Level Response time proposed as 1 60 secs The licensee shall provide the bases for this value and evaluate against this 1 60 secs, and propose as necessary.
06/01/84 55 Revision A
I l
l Item 13:
Station Blackout Item 13:
General The Licensing Basis FSAR, reference 6, page 9.2-10 describes how station blackout causes startup of all Emergency diesel generators and alignment of [NSWS and CCW]. Why is.this not included under this item 13 " Station Blackout."
The Licensing Basis FSAR, reference 7, Section 15.2.9 under LOSS OF QFF-SITE POWER TO THE STATION AUXILIARIES describes a set of Protection Actions for the plant, all which have related response times.
Why is this information not provided under this heading?
The absence of most of.the information on Functional Units and Related Response. ti'mes 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-Driven AFW Pumps" on Station Blackout L
Proposed T.S. response times are 160 secs.
Reference our comment under 7a. and 7b. above.
These values are non-conservative with respect to Regulatory requirements and the licensee shall evaluate and propose.
Item 14:
" Start Motor-Oriven Auxiliary Feedwater Pumps ' on Trip of Main 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.
g Item 15:
Loss of Power:
"4 Kv Emergency Bus Uncervoltage-Grid Degraded Voltage."
Proposed T.S. response time of 111 secs.
Reference our comments under T.S. Table 3.3-3 Item 9 and Table 3.. 4 Item 9 and provide appropriate clarification.
No evaluation is possible at this time.
06/01/84
.56 Revision A
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.
Item [ Foot] Note:
Response time for Motor-Driven Auxiliary Feedwater Pump Starts on All SI signals.
This is proposed as < 60 secs.
Reference our earlier comments f'or its inclusion in' Items 2f., 3f., and 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 propose.
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 items 2g., 2h., 3g., 3h., 4g., and 4i. above.
Notation 2 of this Table states that:
(2) Valves 1XC3058 and 1KC3158 for Unit 1 and Valves 2KC305B 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:
< 30(3)/40(#)
3) 2.b
< 30(
3.b 530(3)/40(4) 4.b Since the functions 2b, 3b and ab are all Reactor Trip functions, please explain.
Since these descriptors are apparently incorrect, provide the correct descriptors.
~.
06/01/84 57 Revision A
o.
Since supercripts (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 Actual overall response time, and has this been incorporated into the Safety Analysis of the Licensing Basis.
The Licensee shall clarify, evaluate and propose.
Lack of accurate information on response times must be considered as non-conservative.
I t
06/01/84 58 Revision A
Section 3/4.4 REACTOR COOLANT SYSTEM Section 3/4.4.1 REACTOR COOLANT LOOPS AND COOLANT CIRCULATION Item:
GENERAL G.1 INTRODUCTION Concerning RCS Operability requirements, in MODE 3-5:
Werefertoourearlierdiscussions$licenseerequirements-andespecially under Section 3/4.1.1, T.S. Page 3/4 1-1, 2 & 2a on Boration Control, T.S.,
Page 3/4 1-20 & 1-21 concerning SHUTDOWN AND CONTROL ROD INSERTION LIMITS and TABLE 3.3-1 REACTOR TRIP SYSTEM INSTRUMENTATION generally, including more particularly items 2-21 (selected) and items 12, 14, 15 and 21.
- Under ou'r 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-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.
As part of this review, and as a safety justification for our concerns, we
^'
require inclusion of the following Occurrences and Considerations in the program, and as early determinants of our proposals in respect of RCS Loop Operability requirements in MODES 3, 4 and 5 (with loops filled).
G.2 DISCUSSION Item:
CONSIDERATION A number of factors determine 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 movable control assemblies and still remain within the unchanged T.S.
condition of the allowable reactivity condition, keff of < 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 1
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 flux trip protection in TABLE 3.3-1 removes all current confidence in "available" Reactor Trips on Neutron Power; the only Safety Related Neutron Flux Trip from zero power subcritical conditions is the Power Range Neutron Flux Low Set Point and the proposed T.S. removes this from operability in MODES 3, 4 and 5.
Further it has a Safety Analysis Limit of 35% power (25% Set Point) and together with related high peaking flux factors under these conditions is sufficient to require all 4 RCPs running to ensure R.C.S. Safety in at least MODE 3.
G.2.1.3 The increased boron concentrations *give less negative and more posi-
'tive moderate coefficients which changes the complexion and. nature of expected responses from " Licensing Bases Events." Under these cir-cumstances, it may not be possible to validly deduce the resulting responses and consequences without related analyses.
G.2.1.4 At this time we see no protection against positive temperature coefficients in MODE 3 [4, 5 & 6].
Proposed T.S. page 3/4 1-4 concerning MODERATOR TEMPERATURE COEFFICIENT requires only that:
"the moderate temperature coefficient (MTC) shall be:
3.1.1.3.b.
Less negative than - 4.1 delta k/k 'F for all the rods withdrawn, 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 should also clarify this T.S. requirement which is 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.
without any Safety Evaluation.
The proposed T.S. are nonconservative with respect to Regulatory Requirements.
They are also nonconservative in respect to cne Licensing Basis.
The Licensee shall evaluate and propose.
G.2.3 In MODE 3, down to P-11, for events initiating Safety Injection, the 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
operability of Reactor Trip from SI in this mode and offers no Safety Evaluation fo.r the proposed change.
Reference our review under Table 3.3-1, Item 17.
The proposed T.S. is not in conformance with the Licensing Sasis, and is nonconservative.
The licensee shall evaluate and propose.
G.2.4 In MODE 3, from P-11, to MODE 5, for events initiating SI, the plant is engineered and can be operated so that only one automatic trip of the reactor may be available; that from containment pressure-high.
On the above bases, plant engineering and operations would not be in conformity with regulatory requirements.
The Licensee shall evaluate and propose.
It may be possible for the plant'to be operated in a manner to conform by not manually blocking the Matn Steam Line Pressure-Low Trip [at P-11] but constraining this blockage to a point at which SG pressure during cooldown is within an acceptable error band of the related Set Point Value.
Under these circumstances, two (2) diverse automatic protections on reactor trip may be available.
In addition the proposed T.S.s do not require operabili.ty of the Reactor Trip /ESF channel in this phas~e of operations below MODE 3
[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 Basis and is nonconservative.
The Licensee shall evaluate and propose.
G.2.5 Diversity of Safety Injection to the maximum extent for related Accident Circumstances can only be retained within existing plant engineering by 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]
i.e., approximately 595 psig in the SG.
The licensee shall evaluate and propose.
G.2.6 EVENTS OF CONCERN (A LIMITED SELECTION)
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 Set 06/01/84 61 Revision A
Point (25%) at a Safety Analysis Limit of 35% (reference page 15.2-3, item 3).
The principal determinant of ultimate power level is Doppler coefficient; contribution of moderator reactivity coefficient is negligible (reference page 15.2-3, items 1 & 2).
The event is 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 OPERABILITY 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 propose.
Furthermore; increased baron concentrations would not change this requirement.
Additional events of a similar nature, with a rapid increase in reactlyity include:
a)
Uncontrolled Boron Dilution (reference 7, pacjes 15.2-13) b)
Startup of an Inactive Reactor Coolant Loop (reference 7, page 15.2-19, revision 7) c)
Excessive Heat Removal Oue to Feedwater System Malfunction (reference 7, page 15.2-30, revision 7) concerning initiation with the reactor at zero power).
ifntil the licensee clarifies availability of MFW during MODES 3 through 5, this must be considered a potential occurrence.
d)
Single rod cluster control assembly withdrawal (reference 7, Page 15.3-9, revision 7).
Although the Licensing Basis is at 100% power, the cir-cumstances from zero power should be reviewed.
e)
Ruoture of a Control Rod Drive Mechanism Housing, at Zero Power (ref-erence 7, Page 15.4-30; revision 42).
f)
Major Rupture of a Main Steam Line (see below).
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 0 212-47d & e, item 25.
Reference 8 proposes that the resulting impact on shutdown margins from this event during MODES 3, 4 ana 5 are imoroved over that of the design basis (of zero power, just critical, Tavg - 557*) as:
" Operating Instructions require that the baron concentration be increased to at least the cold shutdown boron concentration before cooldown is initiated.
This requirement insures a minimum of 1% ak/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
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 conseqt.ences 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 shutdown will be at 1% delta k/k whereas the proposed T.S. margin at Zero Power according to T.S. Page 3/4 1-1 was 1.6 delta k/k.
Please clarify, and at what condition during cooldown the 1.6% 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 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 oc, cur on Power Flux Trip, but on Pressurizer Pressure-Low (SI) (above P-11) [ reference 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.
In 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 less than 4 RCS Loop in MODE 3 under these circumstances would be in noncon-formance with the current Licensing Basis FSAR in a nonconservative manner, and the licensee would be required to 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).
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G.2.6.3 LOSS OF PRIMARY COOLANT: 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 SBLOCA between RCS conditions of 1900 psig and 1000 psig/425*F in Hot Standby, and Q 212-64, 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 Standbye to and including RHR.
The FSAR analysis for SBLOCA in r.eference 7, Section 15.3.1 states that:
"Ouring 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 c.colant p. umps t.hrough the core as they 'are coasting down following trip: 'there-fore upward flow through the core is maintaf' ed."
n 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 7
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 'd 4 loop plants the reactor coolant pumps are assumed 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 Baron (mixing and) Injection to the core.
Rapid baron injection would not be an important issue it boron concentrations are already at cold shut down 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 non-conservative and the licensee must evaluate and propose.
Given the circumstances of the proposed T.S., operability of less than 4 RCS Loops in 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.
l 06/01/84 64 Revision A
= _. _
Additional events of a similar nature to the S8LOCA 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, Power Range Neutron F' lux, High Set Point Trips the Reactor;. Pressurizer Pressure-Low (SI) initiates Safety Injection;
~
reference 7, page 15.4-29',', revision 43, paras. 1 and 5.
Whereas both these protections are proposed. by the T.S. in MODE 2, they are not proposed for MODE 3 which differs from the circumstances of MODE 2 by only a marginal reduction in RCS Temperature.
The FSAR, reference 7, Table 15.4.6-1, revision 42..shows this occurrence as being the only event at Zero Power, analyzed to a smaller N' of RCPs than 4; it has been analyzed for 2 only.
This is an accident with substan-tial but " acceptable to Condition IV occurrences" consequences in terms of fuel cladding damage and RCS overpressurization, but it required at least two RCPs to achieve that (in the Licensing Basis).
Even the two RCPs required in this event are not proposed as being required for MODE 3.
The proposed circumstances in MODE' 3 are clearly non-conservative with respect' to the Licensing Bases.
The licensee shall evaluate and' propose.
Concerning the Large Break " Loss of Coolant Accident."
This is discussed in Accident Analyses in Reference 7, section 15.4.1 for a LOCA from rated power; in Reference 8, item 212.75 page Q 212.47, for a LOCA between RCS conditions of 1900 psig and 1000 psig/425*F in Hot Standbye; in item 212.90(6.3), page 212-61, for a LOCA at and less than 1000 psig/425 in Hot Standbye, and on page Q 212-61b, item 29 for a LOCA in the RHR Mode at 425 psig/350*F.
As for the Small Break LOCA, these analyses are presumably based on a 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 "Concerning Small Break LOCA" which used the W Topical Report WCAP 8356 (reference 19) and which applied to both Large and Small Break LOCAs.
06/01/84 65 Revision A
o.
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 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 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 as zero power.
b)* Start up of an Inactive Reactor Coolant Loop c)
Loss of Offsite Power (reference 7, page 15.2-19, revision 7) 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 power, and not zero power, so that their importance would normally be minimal except 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 positive temperature coefficients in MODE 3 (4, 5 & 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 ir 3rtance 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
O
,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 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 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 events, small and large, down to RCS conditions where RCS flows are not necessary.
The current T.S. are nonconservative in respect to the Licensing Basis in respect to these concerns.
The Licensee shall evaluate and propose.
T.S. SECTION 3/4.4.1:
RCS LOOPS AND COOLANT CIRCULATION 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.
The ACTION S'tatement requires that in the event of loss of 1 [of 4] itCS Loop in MODES 1 & 2, the licensee is required to be in at least HOT STANDBY within 1 br.
The current Safety Analysis Limits in the FSAR, reference 7, page 15.2-16, revision 7, requires an immediate trip of the reactor to RTI & ESFAS response times in the event of loss of 1 RCS pump.
Also, placement of the RCS in Hot Standby with less than one loop operable [without other compensating condi-tions] would be non-conservative in respect of the existing FSAR.
The Action Statement is non-conservative with respect to the current licensing basis and the licensee shall evaluate and propose.
T.S. surveillance requires verification of Reactor Coolant Loco (RCL) circula-tion once every 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
This is unacceptable considering the Safety Analysis limits required above for loss at one pump.
In the event of failure of the Low Reactor Coolant Flow Reactor Trip; the operator should responc 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
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:
" Criterion 22-Protection system independence.
The protection system shall be designed to assure that the effects of natural phenomena, and of 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."
The Facility is non-conservative with respect to this Regulation, the licensee chall evaluate and propose.
This is a generic issue.
The surveillance requirement, every,12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />, is intended to ensure not only that the system is operating, but that it is operating at process conditions 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.
T.S. Page 3/4 4-2:e RCS HOT STANDBY 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 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; and earlier work has shown the Safety Analysis Limits for the p; ant currently requiring at least 4 RCS pumps for this MODE.
The Action Statement allowing 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> with only one RCS loop operable is non-conservative with respect to the current Safety Analysis Limits.
At this time, any No. of loops less than 4 in MODE 3 is non-conservative with respect to the existing FSAR and the plant should be transferred to operation in MODE 4 under these circumstances, with approved maximum normal cooldown rates.
It is recognized there are many protective actions which may provide more flexibility in this MODE within NRC/RCS Safety Criteria but they are not included within the current T.S. proposed by the licensee; further that final ch.ce 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 comoinations of protections which could be proposed may depend on providing the facility with maximum flexibility in other cperations in this MODE 3 consistent with meeting Regula-tory Safety requirement.
See our earlier review under General.
06/01/84 68 Revision A
Given the circumstances of the proposed T.S., operability of less than 4 RCS loops in MODE 3, HOT STAN08Y, 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.
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 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> is also invalid and should be changed.
The surveillance requirement, once every 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />, is intended to ensure not only that the system is operating, but that is is operating at process condi-tions 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.
Survefilece requirements for the S.G. call for a level of 12*[at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. This is not in accordance with the Licensing Basis; this level 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 alternate conditions.
Therefore this exlisting proposal is outside the current 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 (MODE 3):
"*All reactor coolant pumps may be de-energized for up to 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> provided:
(1) no operations are permitted that would cause dilution of the Reactor Coolant System boron concentration, and (2) core outlet temperature is main-tained at least 10*F below saturation temperature."
This is a natural circulation condition; the only Licensing Basis calculation for this is the Natural Circulation calculations of reference 7, page 15.2-27,
" Loss of Offsite Power to Station Auxiliaries"; but at MODE 2 Zerc 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 protections against the possibility of Condition II, III and IV occurrences must involve.
in addition to isolation of all baron dilution sources, securing Reactor Trip System Breakers in the Open Position, closure of MFW isolation ~ valves, isola-tion of MSIVs, and possibly an optimum baron concentration.
At present, the only Licensing Basis for controlling this particular situation is the Emergency Operating Guidelines.
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 F5AR and is non-conservative with respect to that.
The licensee shall provide the reason for this requirement including the expected condition of the Facility, and then analy:e, evaluate and propose.
06/01/84 69 Revision A
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 SHUTDOWN.
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 OPERABLE [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 Coolant Loop A and its associated steam generator [ including related auxiliary feedwater pumps] and reactor coolant pump,*
b.
Reactor Coolant Loop B and its associated steam generator [ including related auxiliary feedwater pumps] and reactor coolant pump.*
~
c.
Reactor Coolant Loop C and its associated steam generator, [ including relating auxiliary feedwater pumps] and reactor coolant pump,*
d.
Reactor Coolant Loop D and its associated steam generator, [ including related auxiliary feedwater pumps] and reactor coolant pump,*
e.
RHR Loop A,*** and f.
RHR Loop B.***
APPLICABILITY: MODE 4.
(Less than 425 psig/350*F]"
The licensee shall evaluate as outlined earlier under Item, General, for RCS loops operability requirements and make proposals relative to the status of many elements of the protection and operations system to ensure that RCS safety is maintained for related Condition II, III and IV occurrences.
At this time, 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
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 separate power divisions to protect against single failure of a bus or distri-bution system.
When the RCS systems are used, the related Auxiliary Feedwater 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, 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 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> are available to the operator to establish an alternate means of core cooling.
This is the time it would take to heat the available RCS 'volum4 from 350*F to the saturation temperature for 400 psi (445 F), assuming the maximum 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> decay heat load.
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 genera-tors from the auxiliary feedwater system.
The time required to establish the alternate means of heat removal is only the fe.w minutes necessary 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 Section 3/4 4.1 Subsection G.2.6.3 "Concerning Large Break Loss of Coolant Accident." See reference 8, page Q 212-47.d where it is described that "After several hours into the cooldown procedure (a minimum time is approximately four hours) when the RCS 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 revi-sion 29 concerning ECCS calculations in a later submittal uncer Revision 23 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 made at 425 psig RCS pressure has been assessed."
The additional Action statement that:
b.
"With no r"eactor coolant or RHR loop in operation, suspend all operatient involving a reduction in baron concentration of the Reactor Coolant System and immediately initiate corrective ACTION to return the required coolant loop to operation."
06/01/84 71 Revision A
and the additional notation that
""*All reactor coolant pumps and RHR pumps may be de-energized for up to 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> provided:
(1) no operations are permitted that would cause dilution of the Reactor Coolant System boron concentration, and (2) core outlet tempera-ture is maintained at least 10*F below saturation temperature."
are unsupportable by present analyses in the FSAR. These proposed T.S.s are the same as for MODE 3 and our relevant comments and requirements under T.S.
Page 3/4 4-2:
RCS HOT STANDBY should be applied to MODE 4.
Emergency Oper-sting Guidelines Apply.
This proposed T.S. is non-conservative with respect to the Licensing Basis.
The ifcensee shall provide the reason for the require-ment including the expected condition of the facility, and then analyze evaluate and propose.
Surveillance, requirement 4.4.1.3.2.should verify S.G. water level at the S,afety Analysis Limit for the Licensir.g Basis, which is the no-load programmed; level, not the current proposed TS vetue which is the S.G. Low-Low Level [ Reactor Trip] and AFW actuation.
Thi, proposed TS is non-conservative with respect to the current Safety Analys7s Limits and the licensee shall evaluate and propose.
Surveillance requirement 4.4.1.3.3 verifying one loop in operation every 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />, is unsupportable as all protective trips on low flow in the RCP loops in this condition have been removed.
If low flow channel trips on the RCP loops are not required to be operable why should the related Alarm be operable.
A low flow alarm for the RHR has been provided by the FSAR under reference 8, page Q 212-56, item:
" Case 1:
The R'eactor Coolant System is closed and pressurized.
The operator would be alerted to the loss of RHR flow by the RHR low flow alarm.
(This alarm has been incorporated into the McGuire design)."
Since currently, these two types of alarms are the only means of alerting the 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 Safety Related and included within the T.S.; and without further analysis at this time, two loons 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 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> is intended to ensure not only that the system is operating, but that it is operating at process conditions which can be evaluated to show that the equipment is capable of performing its
. sign basis Safety Function.
The current surveillance requirements for this item, i.e., for the RCS and RHR systems in Hot Shutdown in T.S. Item 4.4.1.3.3, are absent this information; it is therefore non-conservative and the licensee shall evaluate and propose.
06/01/84 72 Revision A
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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 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 Oump Valves operability is established.
The current T.S. are non-conservative in this matter; they make no provision for operability of this item (see later proposed T.S. page 3/4 7-8a).
[ General comment:
Operability of each of S.G. water level, AFW and ATMOSPHERIC 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.]
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 SHUTOOWN [ MODE 51 WITH LOOPS FILLEO.
~~
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 RHR loop shall be OPERABLE #, or b.
The secondary side water level of at least two steam generators shall be greater than 12P..
The current FSAR requires two (2) OPERABLE RHR trains on two (2) 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 cumps are automatically transferred to a separate emergency diesel power supply.
Therefore; the current licensing basis is that 2 residual heat removal locos shall be operable.
The above provision for either an RHR loop or two steam generators is therefore not in accordance with the Licensing Basis.
The proposed T.S. in this respect is also non-conservative as it would necess-ily require S.G. temperatures greater than 212*F (Atmos Press in SGs) which would place it outside the Cold Shutdown MODE into the Hot Shutdown MODE - which is outside the required Functional MODE.
I The T.S. requirement for one RHR loop in operation and one to be available l
OPERABLE is currently not supportable by analysis evaluating the situation in l
which all RHR cooling is lost in a water solid condition; reference our 06/01/84 73 Revision A
o.
immediately preceeding 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 sufficient time to commence operations of the second RHR loop to the extent necessary to mitigate the consequences of any potential overpressure event in an acceptable manner.
The licensee shall evaluate and propose.
Use of secondary side water level of at least two steam generators is discussed in reference 14 for circumstances in which the RHR is isolated frnm the RCS and its final acceptability for licensing purposes is still not resol /ed.
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 pre'vious 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 of the RHR isolation valve and in which case the RCS returns to MODE 4 with its particular MODE 4 requirements as discussed earlier.
The absence of this as an LCO requirement in the proposed T.S. makes it non-conservative with respect to the Licensing Basis.
The Licensee shall evaluate and propose.
Footnote *:
This item proposes that an only available operational RHR pump may be de-energized for up to 1 hr.
This event has not been evaluated, is not within the Licensing Basis, and is non-conservative.
The licensee should define the circumstances, analyze and evaluate and propose.
The proposed surveillance requirement /4.4.1.4.1.2 prov. ides that "At least one RHR loop shall be determined to be in operation and circulating reactor coolant at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
The items of significance here are Operable Safety Related Flow Alarms with a surveillance frequency ensuring high probability of alarm in the event of an RHR flow failure, and a related concern for overpres-sure protection and recovery.
The licensee shall evaluate and propose.
The surveillance requirement, every 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />, is intended to ensure not only that the system is operating, but that it is operating at process conditions thich can be evaluated to show that 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 snali evaluate and propose.
T.S. page 3/4 4-6.
REACTOR COOLANT SYSTEM - COLD SHUTDOWN, LOOPS ARE NOT FILLED ltem 3.4.1.4.2 requires that:
"3.4.1.4.2 Two residual heat removal (RHR) loops shall be OPERABLE # and at least one RHR loop shall be in operation.*"
Additionally, the current FSAR requires that each of the RHR trains be provided with power from (2) redundant electrical buses so that each pump receives 06/01/84 74 Revision A
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i l
l 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 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> decay heat, air in the r
I 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 oft, if the operator took no action, boiling l
would begin in about five minutes, the water level in the vessel would be down to the level of fuel in about.100 minutes, and the pressure would increase to 550 psi in about 40 minutes (the pressure' rise could be 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 operating RHR systems so that boiling may be eliminated on single failure.
The licensee shall evaluate and propose, f
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 Reactor Vessel flanges and which minimum volume shall be included in the T.S. as an LCO with appropriate surveillance ind 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 attemot to fill and pressurize the reactor coolant system with the centrifugal enarging pumps.
If the system can be pressurized to the range of 400-500 psi, the L
06/01/84 75 Revision A i
i I
l operator could return the plant to heat removsl via the steam generators.
l 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 i
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 Licensfng Basis FSAR, the current T.S. LCOs must be l
l considered non-conservative with respect to the Licensing Basis, and the licensee shall eva,1uate and propose.
Item 4.4.1.4.2, A surveiliance requirement, specifies:
At least one RHR loop shall be determined to be in operation and circulating l
reactor coolant at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
A time delay of 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> is excessive to verify a loop in operation, and this has been considered earlier in this section.
Further the surveillance require-l ment, every 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />, is intended to ensure not only that the system is operating, but that it is operating at process conditions, including instrumentation and control, which can be evaluated to show that the equipment is capable of l
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 Ifcensee shall evaluate and propose.
l Footnote *:
Provides that,
"*The RHR pump may be de-energized for up to 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> provided:
(1) no opera-tions are permitted that would cause dilution of the Reactor Coolant System boron concentration, and (2) core outlet temperature is maintained at least 10*F below saturation temperature."
This departure from the Licensing Basis of two available RHRs with effective i
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 l
100 minutes.
The provision is outside the Licensing Basis in a non-conservative canner and the licensee shall evaluate and propose.
I T/S pace 3/4 4-6(a) Proposed.
A new subsection should be added entitled " REACTOR COOLANT SYSTEM, HOT SHUTDOWN TO REFUELING, APPLICABLE MCDES 4, 5, & 6 which requires a LIMITING CON 0! TION OF OPERATION that two RHR Flow Alarms to Safety Related requirements shall be operable on each RHR loop when only one RHT, loop is in operation under the f
provisions of the Technical Specifications.
Appropriate Action Statements and surveillance requirements shall be applied.
l i
06/01/84 76 Revision A r
L
~ -- _ -. ___._ a _L._ _. _.. ;..
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 SHUTOOWN (MODES 4 and 5)
The T.S. requires that:
"3.4.2.1 A minimum of one pressurizer Code safety valve shall be OPERABLE with a Ifft setting of 2485 psig 2 1%.*
APPLICA8ILITY:
MODES 4 and 5.
AdTION:'
With no pressurizer Code safety valve OPERA 8LE, immediately suspend all operations involving positive reactivity changes and place an OPERA 8LE RHR loop into operation in the shutdown cooling 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 Licensee shall evaluate and propose.
The Action statement is based (reference T.S. page 8 3/4.4-2) on the premise that IN0PERA81LITY 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 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 tne RCS as the plant utilizes two available PORVs on the pressurizer, reset to 400 psig (reference review under T.S. Page 3/4 4 36) in the primary coolant system.
In this respect, the proposed action statement is non-conservative and contrary to the Licensing Basis.
The licensee shall evaluate and propose.
The Surveillance Requirements should contain the minimum disenarge capacity required of this valve as defined in the Licensing Basis.
They should also ensure the maintenance of satisfactory environmental conditions consistent with reliable valve operability.
The licensee shall evaluate and propose.
06/01/e4 77 Revision A
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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:
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 t
6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in at least HCT SHUTDOWN within the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />."
~
Failure of the. Pressurizer Code Safety Valve, in general, would infringe the integrity of the Reactor Coolant Pressure Boundary and the RCS should be brougnt 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 Reactor "Co'olant System" analyzed for the most severe conditions including maximum core power, reference 7, page 15.2-33 revision 7.
This type of event would reouire Emergency Procedures to define the ACTION STATEMENT.
Could other types of failure allow other types of response which could be outside the Emergency Operating Procedures.
The Licensee has not identified others and analyzed and evaluated the related safety to Regulatory Require-ments as a basis for his proposed action.
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 types and appropriate Regulatory requirements including procedures.
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 Action Statement does not recognize these circumstances.
The existing Action Statement is therefore nonconservative with respect to the Licensing Basis; the licensee shall evaluate and propose.
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 a9cessary to satisfy 10 CFR 50 Appendix A, Criterion 20, 32 and 33].
The writer reviewed, in 1983, information pertaining to the GPU/B&W 1awsuit review, and his recollection is that the TMI-2 operators " initially thought that the safety valves had developed a leak in the PORVs 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
,o measurable parameters "in operation" to 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.
Startup of the Inactive Coolant Loop; reference 7 Figure 15.2.6-1, revision 4 Less of Load Accident; reference 7, Figure 15.2.7-5, revision 38 Loss of Normal Feecwater; reference 7, page 15.2-26, revision 7, para. 3 Main Feedwater Line Break Accident, reference 5, Figure 15.4.2.7, 7
revision 38 One Locked Rotor Event; reference 7, Figure 15.4.4-1, revision 32 Safety Valve Operation could also occur on other overpressurization events if same of the early reactor trips fail to operate as expected.
In this matter, the T.S. is nonconservative with respect to Regulatory Reduire-ments.
The Licensee shall evaluate and propose.
This could be a generic issue.
Surveillance Requirements should reference the documents containing the record
~
of the Inservice Testing of the valves for inspection on a regular basis of 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> so that changing operating staff are kept aware of a potentially changing status on a singularly critical item.
T.S. Section 3/4.4.3 PRESSURIZER T.S. Page 3/4 4-9 The APPLICA8ILITY MODES are proposed as 1, 2 and 3.
Item:
Pressurizer Level:
The response of all the analyses of Candition II, III and IV events in refer-ences 7 and 8 depend uoan an initial level of water in the Pressuri:er wnich is programmed as a varying value dependent ucon the Nuclear Power Level.
Addi-tionally, the response 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 so dependent upon this same programmed pressurizer level.
Since therefore this pressurizer level is used in establishing an acceptable outcome of these analyses in terms of the issuance of the operating license, they also represent limiting ea'ditions of operation as defined in 10 CFR 30.J6.
On this basis therefore, the N ensee should provide details of the programmed pressurizer level set points with allowable values consistent with the related channel errors and Safety Analysis Limits used in the FSAR, Section 15 in reference 7.
The licensee shall evaluate and propose.
06/01/84 79 Revision A
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l l
APPLICABILITY MODES:
Pressurizer level should be proposed for MODES 1, 2, 3, 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 so dependent upon this same pressurize pressure.
Since therefore this value of pressurizer pressure is used in establishing an 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 allowable values consistent with the related channel errors and Safety Analysis Limits used -in the Licensing Basis in the FSAR in Section 15 in reference 7, and reference 8.
The licensee shall evaluate and propose.
Appropriate ACTION and SURVEILLANCE procedures should be proposed.
The licensee 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 one of the combination of Block Valve and PORV is INOPERA8LE.
This is a contravention of the regulations which provides under 10 CFR 50.2(v) that:
(v)" Reactor coolant pressure boundary" means all those pressure-containing 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 anc all of the following:
e (i) The outermost containment isolation valve in system piping which penetrates primary reactor containment.
(ii) The second of two valves normally closed during normal re~ actor 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 l
1 06/01/84 80 Revisicn A s
4
.o (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 changes, 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)
S.G. Levels A number of the Accident Analyses in reference 7 depend upon an initial level 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 sucn 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.
b) Steam Generator Pressures Since Steam Generator Pressures and related Saturation Temperatures under 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 orovide the values used as Safety Analysis Limits in related analyses and again respono to the proposition that such values should be included as Set Point and Ailcwacle values as Limiting Conditions of Operation for the facility with appropriate Action Statements.
The proposed T.S. is nonconservative with respect to tne 4
Licensing Basis, by their absence.
c) Please respond to the proposition that this section should also adecuately identify the maximum allowable Steam Generator Pressure under Transient and Accident conditions with appropriate Action Statements.
Maximum SG pressure is one of the Acceptance Criteria for safety.
The current very limited basis for Steam Generator Pressure integrity is completely inadequate.
Please clarify apparent discrepancy between reference 4, Table 5.5.2-1 in whicn the j
steam side design pressure for the Steam Generator is given as 1295 psig and the value quoted in the T.S. Basis Page 3 3/4 7-1 at 1185 psig.
The proposed T.S. is nonconservative with respect to the Licensing Basis, by this absence.
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06/01/84 81 Revision A l
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d) APPLICABILITY MODES 1, 2, 3, and 4:
The current applicability requirements relate to Structural Integrity considerations.
On inclusion of Steam Generator Level and Pressure as determinants of Opera-bility, the ifcensee should evaluate and propose APPLICA8ILITY 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 prop'ose.
T.S. Page 3/4 4-36 (REACTOR COOLANT SYSTEM) OVERPRESSURE PROTECTION SYSTEMS The cu'rrent 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.
l 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:
CA 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:
l (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.
I lt 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 no,nconservative 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 greater than or equal to 4.5 square inches:
We find that this should be confined only to MODE 5, COLD SHUTDOWN,
. 0PS ARE NOT FILLED, and REFUELING OPERATIONS; MODE 6 HIGH WATER LEVEL and MODE 6 LOW WATER LEVEL.
There are no safety analyses to 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
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 shs11 eva?uate and propose.
F i
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06/01/84 83 Revision A
T.S. SECTION 3/4.5 EMERGENCY CORE COOLING SYSTEMS The operability requirements from the McGuire Units 1 & 2 Licensing Basis FSAR are markedly different from those of the W Standard Technical Specifications which have been adepted by the Licensee in his proposed T.S.
The Licensing Basis FSAR requirements are summarized under " General."
General FSAR Reference 8, page Q 212-47, Revision 25, item 212-75, describes the following Operator Instructions and Operator Actions During Shutdown.
"The sequences of events associated with shutdown will be described.
The
. procedures associa.ted with startup will be the same except they.will be in reverse ordsr.
The startup procedures are not presented here to avoid
~
unnecessary duplication.
I Ooerator 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 press,urizer pressure transmitters and from the steamline pressure transmitters.
The SI signal on containment high pressure signal continues to be armed and will actuate safety injec-tion if the setpoint is exceeded.
Manual safety injection actuation is also available.
Also, at 1900 psig, the operator is instructed to close and gag UHI discharge valves.
The UHI hydraulic pump and the gag motors for the UHI isolation valves are de-energized and tagged.
B)
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 l
closed.
II Ooerator Actions Durino Shutdown
.A)
Between 1900 psig and 1000 psig, the ECCS can either be actuated-automatically by the high containment pressure signal or manually by the operator.
06/01/34 83A Revision A i
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B)
Between 1000 psig and 400 psig, a portion of the ECCS can be actuated automatically (containment high pressure signal) or manually by the operator.
The equipment that can be energized are two RHR pumps and one charging pump.
The operator would have to reinstitute power at the motor control centers or switchgear to the remaining safety injection pumps, charging pump, and the accumulator isolation valves.
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 23, 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.
l 2.
At 1000 psig and 425*F, the operator will close and lock out the accumulator isolation valves.
The above plant operating procedures will ensure that the accumulator isolation valves will not be Tocked out prior to about 2-1/2 hours after reactor shutdown for a cooldown rate of 50 F/hr.
(
A conservative analysis has defined that the peak clad temperature resulting from a large break LOCA would be significantly less than the 2200 F Acceptance Criteria limit using the ECCS equipment available l
2-1/2 hours after reactor shutdown.
The following assumptions were used in the analysis:
1.
The RCS fluid is isothermal at a temperature of a25*F and a pressure of 1000 psig.
2.
The core and metal sensible heat above 425 F has been removed.
3.
The hot spot occurs at the core midplane.
4.
The peak fuel heat generation during full power operation of 12.88 kW/ft j
(102% of 12.63 kW/ft) will be used to calculate adiabatic heatuo.
5.
At 2-1/2 hours decay heat in conformance with Appendix K of 10 CFR 50, the peak heat generation rate is 0.179 kW/ft.
06/01/84 84 Revision A
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6.
Two low head safety injection pumps and one high head charging pump 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 reactor vessel at the end of blowdown.
8.
A large cold leg break is considered.
For a postulated LOCA at the cooldown condition of 1000 psig, previous calculations show that the clad does not heat up above its initial temperature during blowdown.
Proceeding from the end of blowdown and assuming adiabatic 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 hot spot to the steam boiloff and entrained water will commence.
This 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 S1 pumps have completely filled the downcomer.
No cooling due to steam boiloff is considered at the hot spot until the downcomer is filled.
The full downcomer situation may then be compared with the results of the ECCS analysis in the SAR to obtain a bounding clad 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 wnen 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 icwer core power level in effect; decay heat input at any given elevation is less in the shutdown case.
The combination of more rapid reficoding 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.
06/01/84 85 Revision A
4 Repeating the above calculation assuming the loss of a low head safety infection pump yields clad temperature of 1653*F and 1760*F for Cases 1 and 2, respectively.
These results provide additional assurance that the peak clad. temperature will not exceed 2200*F because, as stated above, in the shutdown case more rapid reflooding and lower power in the fuel insures that the clad temperature rise during reflood wil be less than for the design basis case.
Based upon the analysis as presented above, it can be concluded that in the unlikely event of a LOCA at shutdown conditions, the peak clad
[
temperature will be less limiting than that of the design base calculation.
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 four hours after reactor shutdown.
The integrity of the core after a postulated LCOA is assured if the top of the core remains covered by the resultant two phase mixture.
A conservative indication of time available for operator action is obtained by calculating the time required for the top of the core to just uncover.
A calculation has been performed to confirm that margin for operator action does exist to prevent core uncovery.
This conclusion persists even under an assumption of ten minute delay for operator reaction time.
Assumptions:
(a) The system pressure essentially reaches equilibrium with containment by the time the volume of water above the bottom of the hot. legs is removed.
(b) Upper plenum fluid volume between the top of the core and bottom of hot legs is the only upper plenum 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.
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, tne total liquid mass in the downcomer, core, and upper plenum was calculated, i.e., a mass-initial condition.
Again by hydrostatic pressure balance, the height of liquid in the downcomer when the top of the core is just about to uncover was calculated.
This information along with core volume is used.a develop a mass-final condition.
That is, the mass is 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.
-06/01/84 86 Revision A 1
i
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 which 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 when 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; enabling 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 hatter th.a that in the reactor coolant system.
The applicant has shown that in neither case was 10 CFR Part 50,
(
Appendix G limit reached.
For the latter case (tnat 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 greater than 50 degrees Fahrenheit.
The staff requires that the technical specifications require that the reactor coolant system may not be cooled to temperatures lower than 300 degrees Fahren-
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hei. without the overpressure protection system enabled, and unless both 06/01/84 87 Revision A
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power-operated relief valve trains are operable, in order to assure suitable overpressure protection for the reactor coolant system when water-solid.
In addition, the technical specifications will state that the temperature of the fliuid in the secondary side of the steam generator will not exceed the temp-crature of the fluid in the reactor coolant system by greater than 50 degrees Fahrenheit when the reactor coolant system fluid temperature is less than 300 degrees Fahrenheit since the applicant's calculations did not assume differences greater than 50 degrees Fahrenheit.
The applicant provided data to show that the power-operated relief valve opens within the time specified in the analyses.
The system meets the single failu.e criteria as only one of the two trains is required for overpressure mitigation. Means ar.e provided to test and calibrate the system.
It has been designed in accordance with the Institute of Electrical and Electronics Engineers Standard 279-1971, " Criteria for Protecti.on Systems."
This system meets the staff requirements for an overpressure protection system with the reactor coolant system water-solid and is acceptable. We consider this matter resolved.
The required status of the ECCS systems required by the existing Licensing Basis FSAR are briefly summarized:
Above 1900 psig (in MODES 1, 2, and 3):
All ECCS systems are OPERABLE.
Between 1900 psig and 1000 psig/425 F; upper head injection isolation valves area closed and gagged, de-energized and tagged.
Between 1000 psig/425 F and 425 psig/350 F (in MODE 3):
Upper head injection isolation valves remain closed and gagged and de-energized; cold leg accumulator isolation valves are closed and breakers racked out,1 centrifugal and 1 reciprocating charging pump and 2 safety injection pumps are isolated, and rendered inoperable by 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 (as for the preceding phase of MODE 3) with the exception that remaining equipment is re-aligned for RHR operation with the capability of re-alignment to ECCS.
[UHI, Cold Leg Accumulators, 1 cent. CP & 1 Recip. CP, and 2 SI pumps are effectively electrically isolated.] RHR PORVs are rendered ocerable during water solid operation, below 300 F.
These requirements are substantially different from those of the W STS which the licensee has adopted for his facility contrary to his Licensing Basis as disclosed in the FSAR and SER to the above references.
T.S. SECTION 3/4 5.1 ACCUMULATORS / COLD LEG INJECTION Item:
APPLICABILITY MODE 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 j
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 under " 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 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 set point of approx. 450 psig.
The speci.fied 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 unecessarily 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 within 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
and (2 of 2) revision 35, " Interlocks for ESFAS," nor in the related Logic Diagrams.
The LCOs of the Licensing Basis FSAR require that this Cold Leg Infection Accumulator be made 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 shall 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 (Procosed)
An additional T.S. Section is-proposed that pr'ovides: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."
Appropriata 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 of this specific provision makes the proposed T.S. non-conservative.
The licensee shall evaluate and propose.
T.S. Page 3/4 5-3.
UPPER HEAD INJECTION -
ltem:
APPLICABILITY MODE.
1 The Applicability Mode given as MODES 1, 2, and 3* where
- signifies Pressurizer 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, i
thougn 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.
Frem his detailed analysis, the licensee should evaluate and propose a lower limit i
to this temperature condition of >425 F.
l Etem 3.5.1.2.c Nitrogen cover pressure is specified as between 1206 and 1254 psig.
The Licensing Basis FSAR, reference 29, page (1 of 5), revision 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 shail evaluate and propose.
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 Licen'ing Basis.
The licensee shall evaluate and propose.
s Action Items a & b require HOT STAN0BY, 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 eva10 ate 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)
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, wouTd 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 Proceoures 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 Basis.
The licensee shall evaluate and propose.
T.S. Section 3/4.5.2 ECC SUBSYSTEMS -Tava 2 350 F The title should be amended to read as:
ECCS SUBSYSTEMS - PRESSURIZER PRESSURE > 1000 psig/RCS Tavg2 25 F 4
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 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 advise how they might meet minimum flow conditions to intact loops dating Accident Occurrences.
T.S. Section 3/4.5.2.A Proposed A proposed new Section which would be titled:
ECCS Subsystem - Applicability between 1000 psig/425*F and 425 psig/350*F.
. This would provide for:
One ECCS subsystem comprising the following shall be OPERABLE:
a.
One OPERABLE centrifugal charging pump,#
i b.
One OPERABLE RHR heat exchanger, c.
An OPERABLE flow path.
Also,' one ECCS subsystem comprising the followin'g shall also be OPERABLE b.
One OPERABLE RHR heat exchanger, c.
An OPERABLE flow path All breakers for all safety injection pumps and all but the one operable centrifugal charging pump are opened, locked and tagged (reference earlier information).
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 (LTOP).
The proposed T.S. requires all CI and SI pumps to be available during these conditons and is therefore non-conservative with respect to the Licensing Basis and particularly in respect.
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. Sectior 3/4.5.3 ECCS Subsystem - Tavo & 350*F
.This title should be amended to read ECCS Subsystems - 425 psig/350 F to COLD SHUTDOWN 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
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 T,itle 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 comprising One Operable Centrifugal Charging Pump (CCP)'
a.
b.
One Operable RHR Heat Exchanger c.
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 oretection 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 OVEPRESSURE 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 li.censee shall evaluate and propose.
The LCOs of the Licensing Basis FSAR require the same operability of ECOS equipment as is required for TS 3/4 5.2A Preposed.
So that in addition to:
06/01/84 93 Revision A
One ECCS subsystem comprising the following shall be OPERABLE:
a.
One OPERABLE centrifugal charging pump, b.
One OPERABLE RHR heat exchanger, c.
An OPERABLE flow path which is the same as for the proposed T.S., it is also required that:
One ECCS subsystem comprising the following shall also be OPERABLE:
b.
One OPERABLE RHR heat exchanger,
~
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.
(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.
m Additionally the Licensing Basis requires that ech of these subsystems be independent and receive power from two (2) redundant Emergency Buses and Power Sources.
The absence of any such provision in the proposed T.S. makes it non-conservative with respect to the Licensing Basis.
The Licensee J
shall evaluate and propose.
T/S Section 3/4.5.4 BORON INJECTION SYSTEM / BORON INJECTION TANK.
Item:
APPLICABILTY 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 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 Mcde 4 must be considered non-conservative.
The licensee should evaluate and propose.
Ztem:
The ACTION Statement should be clarified to include [
] that in the event of inoperablity of the BIT tank, the RCS be borated to [a baron concentra-tion which will give] a SHUTDOWN margin of 1% delta k/k at 200 F.
The li..nsee shall clearly indicate, that this item is not 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 SHUTD0%N overly conservative; if minimum volumetric requirements are necessary, can 06/01/84 94 Revision A
]
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:
APPLICA8LITY MODES 1, 2, 3, 4.
The current MODES 1, 2, 3 and 4 which includes an LCO for 372,100 gallons must be extended to MODE 5 and MODE 6 (limited) to meet the FSAR requirements in reference 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 valve is resolved by the expected Operability of the RWST providing 5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> of injection flow.
The recovery description also means that the RWST must be available in MODE 6 until the vessel head is removed a'nd the refuelincj canal is filled to its speci'fied level.
It must also be available at termination of core alterations - in Mode 6, when drainage of the refueling canal commences until the Reactor Vessel Head is tensioned, when the RCS then moves into MODE 5.
The proposed T.S. is non-conservative with respect to the Licensing Basis.
The licensee shall evaluate and propose.
Actiori Statement:
The proposed ACTION should be modified [
] as follows:
l With the RWST Inoperable, restore the tank to OPERABLE status within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />, or be in at least HOT STAN0BY [and borated to a baron concentration which will give a shut down margin of 1". delta k/k at 200*F and a minimum of 2000 ppm]
within [the next] 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.
The Licensing Basis FSAR requires Safety Injection of 2000 ppm Boron to mitigate the nuclear power consequences of any accidents which may initiate during this period; if the RWST is not available, then Baron Concentration in the RCS snould 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 SHUTC0hN is required.for each INOPERABILITY REQUIREMENT, or whether alternate mitigating Actions are,possible.
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 stean generator shall be OPERABLE with lift settings as spe(fied in Table 3.7-3.
APPLICABILITY: MODES 1, 2, and 3.
ACTION:
With four reactor coolant loops and associated steam generators in a.
operation and with one or more main steam line code safety valves inoperable, operation in MODES 1, 2, and 3 may proceed pro'vided, that within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />, 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 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.
-b.
With three reactor coolant loops and associated steam generators in operation and with one*or more main steam line code safety valves associated with an operating loop inoperable, operation in MODES 1, 2, and 3 may proceed provided, that within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />, either the inoperable valve is restored to OPERABLE status or the Power Range Neutron Flux High Trip Setpoint is reduced per Table 3.7-2; otherwise, be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTD0'nN with the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.
Our concerns in this section are parallel to those in our review under T.S.
Section 3/4.4.2 SAFETY VALVES.
Failure of Steam Generator Code Safety Valves infringe basic safety criteria for Reactor Protection through its imoact on SG/RCS system response unoer Condition II, III, and IV occurrences.
It also affects the integrity of the Primary Containment Boundary.
We do not find an adequate consideration of the alternate type of Safety Valve
~
Failure that can occur, and their related significance, uoan 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 tr 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 coerable in MODES 1, 2, 3, and 4, with a further requirement i
06/01/84 96 Revision A
e 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 4 and 5 in accordance with our review under Table 3.3-3 ESFAS INSTRUME!!-
TATION, Items 7 a, b, c, d, e, and f.
The conclusions from that review are:
The proposed T.S. items are generally non-conservative with respect to the Licensing Basis. The licensee shall evaluate and propose.
Item 3.7.1.2.b.
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 justiff-cation has been provided:
Reference 15, Recomm'endation GL-3, requires the Steam-Turbine AFW pump.in the event of complete loss of AC power for a perio'd of 2 hrs and beyond.
This will require operability down to the lowest pres-sures 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 from MODE 5) to cover licensing require-ments 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 Driven Feedwater Pumps:
They are supplied with steam from two steam generators from main a.
steam lines after the flow restriction orifices at outlets from the Steam Generators.
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 speciff-cations which will ensure operability of the Turbine-Driven 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 Licen.ing Basis.
In his evaluation, the licensee should advise if Item le of Table 3.3-5 ESFAS s
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
Condition III and especially Condition IV Occurrences of the Licensing Basis, and if the existing Accident Analyses remain valid.
Item 4.7.1.2:
SURVEILLANCE REQUIREMENTS The Technical Specifications, page T.S. 3/4 7-4 requires each motor driven (MD)
AFW pump to supply 450 pgm at greater than or equal to 1210 psig.
This is at entrance to the Steam Generators according to the T.S. Basis on T.S.
page B 3/4 7-2.
However, we note that the FSAR Accident Evaluation; reference 7, section 15.4.2.2.2, and the description of the AFW system in reference 5, refer to a
, total supply of 450 gpm from MDAFW pumps to three intact steam generators.
Further, this is parallel with a description in the Accident Analysis on 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 ta 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 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, ana 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 thit the required flow division is maintained'in an acceptable manner.
At this j
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. Pace 3/4 7-5c Procosed:
CONDENSATE STORAGE TANK SYSTEMS It is proposed that a new item be added to the Technical Specificaticns 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
~
.=
=
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 storage tank inoperable, within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> either a.
Restore the CST to OPERABLE status or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in HOT SHUTDOWN within the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />, or Demonstrate the OPERABILITY of the Nuclear Service Water System and Standby Nuclear Source Water Pond (alternate water source) as a backup supply, and align to the auxiliary feedwater pumps, and restore the condensate storage tank to OPERABLE status within 7 days, or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in HOT SHUTDOWN within th following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.
SURVEILLANCE REQUIREMENTS should include
/
a.
The condensate storage tank system shall be demonstrated OPERABLE at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> by appropriate measures when the tank is the supply source for the auxiliary feedwater pumps.
b.
The Nuclear Service Water System and Standby Nuclear Source Water Pond shall be demonstrated OPERABLE at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> by appropriate measures.
Additionally, an evaluation of and provision will need to be made concerning potential loss of AFW supplies during loss of suction and change-over to alternate AFW sources.
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 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 mace 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.
T.S. Pace 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
4
'The requirements within the Licensing Basis for Main Steam Line Isolation are discussed in this review under Table 3.3-4, Item 4.
The Licesing Basis does require operability 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" and the Licensing Basis FSAR under reference 4 Table 6.2.4-1 (sheet 7 of 11)
Revision 4 and that Primary Containment Integrity is required in MODES 1, 2, 3, and 4 according to proposed T.S. Section 3/4.6.1, T.S. Page 3/4 6-1.
The proposed T.S. is non-conservative with respect to the Licensing Basis; the Licensee shall evaluate and propose.
T.S. Page 3/4 7-8a Proposed:
STEAM GENERATOR POWER O'PERATED RELIEF VALVES (SG PORVs)
The proposed T.S. does not include these valves which are required to enable the plant to be cooled down under natural circulation conditions (under Loss of Offiste Power].
The Lir.:msing 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 Ztem 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 Values for % Rated Thermal Power DURING START 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 i
ovent 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 45% (of full load main steam flow to atmosphere) described under reference 22, page 10.1-2, revision 8, para 3, if they can be available under Loss of Offsite Power.
An appropriate Action Statement should be provided.
If the additional 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 l
06/01/84 100 Revision A
._~
o to the SG PORVs is contrary to Regulatory Requirements which have been excluded from the Licensing Basis.
The Licensee shall evaluate and propose.
T.S. Section 3/4.7.3:
COMPONENT COOLING WATER SYSTEM The proposed T.S. requires that:
3.7.3 At least two independent component cooling water loops shall be OPERABLE.
APPLICABILITY:
MODES 1, 2, 3, 4 ACTION:
With only one component cool'ing water loop OPERABLE, restore at least two loops to OPERABLE status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.
The SER for the plant under reference 10, summarizes the fo'11owing Licensing Basis for the Component Cooling System:
9.2.4 Comoonent Cooling System The component cooling system provides cooling water to selected nuclear auxiliary components during normal plant operation and cooling water to safety-related systems during postulated accidents.
The component cooling system is designed to:
(1) remove residual and sensible heat from the reactor coolant system via the residual heat
~
removal system during shutdown; (2) cool the letdown flow to the chemical and volume control system during power operation; (3) cool the spent fuel pool water; and (4) provide cooling to dissipate waste heat from various primary station components during normal operation and postulated accident conditions.
Active system components necessary for safe plant shutdown are designed to include at least 100 percent redundancy.
The ccmponent 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 equipment.
The component cooling water pumps are powered by recundant emergency buses.
The portion of the component cooling water system serving the residual heat removal system meets the single failure crite 'on for active components.
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 m
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 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 and testing.
We conclude that the component cooling water system is acceptable.
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.
The proposed T.S. completely ignores, without any evaluation, the Licensing Basis requirement for this system in MODES 5 & 6.
The current T.S. are non-conservative with respect to the Licensing 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 MODES 5 and 6.
dithin 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 MCDES 5 and 6.
The proposed T.S. is nonconservative with respect to the Licensing Sasis.
The licensee shall evaluate and propose.
T.S. Section 3/4.7.5 STANDBY 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
....in the intake structure" 06/01/S4 102 Revisien A
o 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-vided little information on the range of AFW temperatures used in his analyses 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.a.ny 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 near freezing temperatures must be considered as nonconservative in respect of certain events.
The Licensee shall evaluate and propose.
APPLICA8LE MODES:
The system is required in all MODES 1, 2, 3, 4, 5, & 6 to handle heat rejection re~quirements 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 and propose.
Reference 6, page 9.2-13, revision 39, states that "In the event of solid layer of ice" forms on the SNSWPg the operating train (of the Nuclear Service Water [NSW] system] is manually aligned to the SNSWP.
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 rea. son, survei Pance requirements ensuring this action snould 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.
06/01/34 103 Revision A
l i
i T.S. Section 3/4.9 REFUELING OPERATIONS l
l' l
T.S. Item 3/4 9.1 BORON CONCENTRATION Additional LCOs are necessary to meet the requirements of reference 8, page 15.2 - 14, revision 10 concerning Accident Evaluation for Section 15.2.4, Uncontrolled Baron Oilution.
The boron dilution analyses of this reference 7, j
provides that, during refueling:
"A minimum water volume in the Reactor Coolant System is considered.
a.
This corresponds to the volume necessary to* fill the reactor vessel l
above the nozzles to ensure mixing via the residual heat removal l
loop."
'b.
Neutron sources are' installed in the core and the sou'rce range 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 l
INVFE 5630) is active providing an alarm to the operator when the
(
flow rate from the charging pumps exceeds 175 gpm.
I 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
~
the loop provide ' acceptable flow, including NPSH conditions, at inlet to the RHR pumps.-
These conditions are appropriate LCO's to 10 CFR 50.36; their current absence from the T.S. for this MODE is a non-conservative situation in respect of the Licensing Basis, and the Licensee shall evaluate and propose.
The current SER, Supplement No.1, reference 11,15-1, provides that:
i "During refueling the applicant has committed to isolate all sources of unborated water connected to the primary system refueling / canal / spent fuel.
l Be do note that Surveillance Requirement T.S. 4.9.1.3 does provide for verifying l
that valve No. INV-250 is closed, under administrative control in support of l
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 i
locked under the proposed administrative control.
Also notice, that reference j
7, page 15.2 - 14, revision 10 states that:
"The other two paths are through 2 inch lines, one of wnich leads to the volume control tank with the other bypassing this tank.
These lines contain flow control valves INV171A and INV175A respectively."
l l
06/01/34 104 Revision A
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 specificationofthepositionofmovablec8Nrolassemblies.
We also note the need to add, according to reference 7, page 15.2-14, revision 10, that the baron concentration is to give a shutdown margin of at least 5 per cent delta k with all the rod cluster control assemblies out.
The additional requirement underlined should be a part of the LCO 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 licensee shall evaluate and propose.
I.n 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 available.
This proposal is therefore outside the LCO for the FSAR in a non-conservative manner.
The Licensae shall evaluate and propose In his Basis, on T.S. Page 3/4 '9-2, 18i t.,r., the licensee has proposed that:
"With the reactor vessel head g,cve. %d 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!n hours for the available water inventory to boil.
In that case, a number of alternates are proposed to resolve the situation and almost invariably, electric power is required, and in most cases the RHR equipment is used.
If the basis for the licensee's request here is to enable him to operate 06/01/84 105 Revision A
8-with only one available electrical bus, it is unacceptable, as the loss of one operable RHR on loss of the only available electrical bus, with containment isolation required in 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />, 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 separate 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 locps ta such as to provide acceptable flow, including NPSH conditions, at inlet to the RHR pumps.
Absent those required conditions from the Limiting Conditions of Operation could make them non-conservative.
The licensee shall evaluate and propose.
The ACTION STATENENT provides that with no RHR loop operable, the containment should be closed within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.' Information in reference.8, page Q 212.
under Case 2 sho'ws that if RHR-is absent [by isolation of the RCS/RHR inlet valve] that:
"Approximately 2.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> are available to the operator to estab.lish an i
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 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> decay heat load."
The current value of 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> appears less conservative than this calculated value of 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> within the FSAR.
The licensee shall evaluate and propose.
The current surveillance requirement:
4.9.8.1 "At least one RHR loop shall be verified to be in operation and circulating reactor coolant at a flow rate of greater than or equal to 3000 gpm at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />."
is 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 uo to 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> per 8-hour period during the performance of CORE ALTERATIONS in the vicinity of the reactor vessel hot legs."
The licensee shall provide the basis for this proposal including safety svaluation, 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 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> down to lh hours, and for a considerable period of time thereafter whilst temperatures are 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 Review of available responses to the consequences of a fail closed RCR/RHR isolation valve, include many procedures using the containment sump.
To allow 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 The subtitle should read as 3/4.9.9 HIGH WATER LEVEL Clarify.by addition of the term HIGH
~
T/S Page 3/4 9-11 REFUEL"ING OPERATIONS LOW WATER LEVEL APPLICABILITY:
MODE 6 when the water level above the top of the reactor vessel flange is less than 23 feet.
GENERAL REVIEW:
Whereas the existing FSAR under reference 20, page 5.1-7 discusses Refueling, it does not provide for a sustained period of normal operations under these Low Water Level conditions.
The FSAR provides that:
" Refueling Before removing the reactor vessel head for refueling, the system temperature has been reduced to 140*F or less and hydrogen and fission i
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 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 develooed for these circumstances.
However analyses have been undertaken for the water inventories and temperatures in the RCS system that might apply under those conditions.
Presumably therefore, the "0PERATING MODE - LOW LEVEL" is a long term changing condition following Cold Shutdown, with loops drained and bolts tensioned changing to bolts untensioned and removal of the head, as concomitant i
flooding of the reactor vessel cavity continues.
At this time therefore, we cannot presume that the consequences of the case of single failure of the RHR/RCS isolation valve used as Case 3 in FSAR reference 8, page Q21-57, does not also apply under this MODE. We will use these consequences to evaluate.
Further, since this is effectively a long term changing condition, in the FSAR, it is not acceptable to allow some of the provisions requested such as one hour for the performance of CORE ALTERATIONS--which by T.S 3/4 9.9 are only permissible under that specification with at least 23 feet of water over the reactor vessel flange.
06/01/84 107 Revision A
It is proposed that an additional item be added to the current statement of APPLICABILITY to the effect that:
This MODE shall not to be used for continuous normal operations, but only as a set of circumstances occurring during the 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:
"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; re.ference 20, page 5.5-24, revision 9.. Without
,this requirement, the T.S i.s less conservative than the FSAR and the license.e shall evaluate and propose.
Additionally, the current FSAR, reference 8, page Q212-57, revision 25, describes that in the event of loss of flow caused by closure of the RHR/RCS isolation valve, [and also by cessation of flow in the system]
"The. operator would.be alerted to.the loss of RHR ficw by the RHR low flow alarm.
Assuming worst case conditions (maximum 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> decay heat,--and the RCS drained to just below the vessel flange) and making conservative assumptions about the amount of water available to heat up and boil 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 operating system so that the operator can respond within 10 minutes to commence 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.
Accitionally, the above information defines an LCO of a minimum volume of water for the related event in which the RCS is drainea 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
0,
Footnote *:
provides.that,
"* Prior to initial criticality the RHR loop may be removed from opera-tion for up to 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> per 8-hour period 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 per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />."
A time delay of 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> is excessive to verify a loop in operation, and this has been considered earlier in this section.
Further, the surveillance requirement, every 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />, is intended to ensure not only that the system is operating, but that it is operating at process conditions, including instrumentation and control, which can be evaluated to show that the equipment is capable of performing its Licensing Basis safety function.
The current requirements for this item are absent most of this information; it is therefore non-conservative and the licensee shall evaluate and propose.
The current ACTION STATEMENT calls for containment closure in 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> [i.e.
240 mins].
Earlier conservative calculations for this MODE show that loss of all RHR in this MODE can cause boiling in 5 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.
I l
06/01/84 109 Revision A i
Addenda T.S. SECTION 3/4.5 EMERGENCY CORE COOLING SYSTEMS T.S SECTION 3/4.4.4.1 RCS LOOPS AND COOLANT CIRCULATION / HOT SHUTDOWN MODE 4 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, 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 acceptable surety that:
a)
The Reactor Coolant Pressufe. Boundary (RCPB) valves on the RHR/RCS suction line are confirmed closed in MODES 1, 2, & 3.
b)
.ThattheRCP8valvesintheRHR/RCSsuctionlineareindividually 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 openi.ng of valves to the RWST in accordance with 10 CFR 50 App A, Criterion 20 [with appro-priate provision,for RHR pump protection].
The current position in respect of c above appears to be absent those requirements and therefore non-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.
Proposed T/S Page 3/4 5-6, item 4.5.2.d,1) b) appears incorrect:
it provides that, in establishing ECCS operability:
d.
At least once per 18 months by:
1)
Verifying automatic isolation and interlock action of the RHR System from the Reactor Coolant System by ensuring that:
a)
With a simulated or actual Reactor Coolant System pressure signal greater than or equal to 425 psig the 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 grecter than 475 psig, the 06/01/84 110 Revision A
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.d e'45 T o w h non -CorreOrd/4.
A/doyhe 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 va ge of 560 unless there are Set Point and Channel considerations - The pressure is less conser-vative than the Licensing Basis FSAR value.
o 06/01/84 11' Revision A
^
LIST OF REFERENCES 1.
Letter from H. 8. Tucker (0.P.Co) to H. R. Denton (NRC) dated September 27, 1982 to the subject of "McGuire Nuclear Station."
2.
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."
-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, Duke Power Company, McGuire Nuclear Station, Units 1 and 2, Rev. 45.
5.
U.S.. Nuclear R,egulatory Commission, Final Safety Analysis Report, Volume 7, 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, Duke Power Company, McGuire Nuclear Station, Units 1 and 2, Rev. 45.
9.
Deleted 10.
U.S. Nuclear Regulatory Commission; Office of N' clear Reactor Regulation; u
" Safety Evaluation Report; McGuire Nuclear Station Units 1 and 2, Duke Power Company," NUREG-0422, on Occket Nos. 50-369 and 50-370, March 1, 1978.
11.
U.S. Nuclear Regulatory Commission, Office of Nuclear Reactor Regulation,
" Safety Evaluation Report, McGuire Nuclear Station Units 1 and 2, Duke Power Company," NUREG-0422, Supp. 1, on Occket Nos. 50-369 and 50-370, May 1978.
12.
U.S. Nuclear Regulatory Commission, Office of Nuclear Reactor Regulation,
" Safety Evaluation Report, McGuire Nuclear Station, Units 1 and 2, Duke Power Comoany," 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 Compat,," 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
15.
U.S. Nuclear Regulatory Commission, Office of Nuclear Reactor Regulation,
" Safety Evaluation Report, McGuire Nuclear Station Units 1 and 2, Duke power Company," NUREG-0422, Supp. No. 5, on Docket Nos. 50-369 and 50-370, April 1981.
16.
Memo from R. W. Houston to T. M. Novak on the subject of " Staff Review and Input to SER Supplement No. 6 for McGuire Nuclear Station Units 1 and 2".
Dated February 08, 1983.
17.
Letter from H. B. Tucker (D.P.Co) to H. R. Denton (NRC) on the subject of McGuire Nuclear Station, Units 1 and 2, filing amendment No. 71 to its Application for License for the McGuire Nuclear Station and Submitting Revision 45 to the Final' Safety Analysis. Report.
Dated February 16, 1983.
18.
Letter from W. O. Parker (D.P.Co) to H. R. Denton (NRC), dated Oct. 8, 198.1 on the subject of McGuire Nuc. lear Station, Unit 1 and submitting copies 'of Report identified as " Westinghouse Reactor Protection System /
Enginee'ed Safety Features Actuation System Setpoint Methodology, Duke r
Power Company, McGuire Unit 1," by C. R. Tuley et al. and dated April 1981, published by Westingnouse Electric, Nuclear Energy Systems, PROPRIETARY.
19.
Westinghouse Electric Ccrporation, PWR Systems Division " Westinghouse Emergency Core Cooling System - Plant sensitivity studies, WCAP-8356.
August 1,1974.
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 (0.P.Co), dated May 17, 1983 on the subject of OL Condition 2.C.(11)g, Anticipatory Reactor Trip (II.K.3.10) (McGuire Nuclear Station, Unit 1)'.
22.
U.S. Nuclear Regulatory Commission, Final Safety Analysis Report, Volume 9, Duke Power Company, McGuire Nuclear Station, Units 1 and 2, Rev. 45.
t 23.
Letter from W. O. Parker (D.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 McGaire Nuclear Station, Unit 1. Docket No. 50-369, License No. NPF-9 Startup Report, February 15, 1982.
26.
Memo for RSB, CPB, ICSB Members from Brian W. Sheron (RSB), Carl H.
Berlinger (CPB), Faust Ross (ICSB) dated April 12, 1983 on the Subject f-of Inadvertent Baron Dilution Events.
l 27.
Westinghouse Electric Corporation, Nuclear Energy Systems Topical Report, l
Overpressure Protection for Westinghouse Pressurized Water Reactors, WCAP-7769, Rev. 1, June 1972.
t t
06/01/84 113 Revision A
,o.
- 28. Westinghouse Electric Corporation for the Westinghouse Owners 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.
e6 4
06/01/84 114 Revision A
TABLE 1 SECTIONS REVIEWED BY REACTOR SYSTEMS BRANCH SECTION PAGE 2.1 SAFETY LIMITS 2.1.1 REACTOR CORE...................................................
2-1 2.1.2 REACTOR COO LANT SYSTEM PRESSURE................................
2-1 FIGURE 2.1-1 REACTOR CORE SAFETY LIMIT - FOUR LOOPS IN OPERATION.....
2-2 2.2 LIMITING SAFETY SYSTEM SETTINGS 2.2.1 REACTOR TRIP SYSTEM INSTRUMENTATION SETPOINTS...................
2-4 TABLE 2.2-1 REACTOR TRIP, SYSTEM INSTRUMENTATION TRIP SETPOINTS.......
2-5 3/4.0 APPLICABILITY.................................................
3/4 0-1 3.4.1 REACTIVITY CONTROL SYSTEMS 3/4.1.1 80 RATION CONTROL Shutdown' Margin - T,yg > P'rogrammed No Load T,yg........... 3/4 1-1 Shutdown Margin - T
< Programmed No Load T and>200*F..ayg,,,,,,,,,,,,,,,,,,,,,,,yg,,,,,,,,,,,,,
a S hutdown Margi n - T,yg < 200*F.............................
3/4 1-3 Moderate Temperature Coefficient............................ 3/4 1-4 Mi nimum Temperature for Criticality........................
3/4 1-6 3/4.1.2 80 RATION SYSTEMS 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 Charging Pump - Standbye, Shutdown and Refueling...........
3/4 1-9 Chargi ng Pumps - Operati ng.................................
3/4 1-10 Borated Water Sources - Shutdown...........................
3/4 1-11 Borated Water Sources - Operating..........................
3/4 1-12 Instrumentation............................................
3/4 1-13a 1
06/01/84 115 Revision A
SECTION PAGE TABLE 3.1-1 ACCIDENT ANALYSES REQUIRING REEVALUATION IN THE EVENT OF AN INOPERABLE FULL-' LENGTH R00......................
3/4 1-16 Position Indication Systems - Operating....................
3/4 1-17 Position Indication System - Shutdown......................
3/4 1-18 Rod Drop Time (Units 1 and 2)..............................
3/4 1-19 Shutdown Rod Insertion Limit (MODES 1 & 2).......,..........
3/4 1-20 Shudown Rod Insertion Limits (Modes 3 - S).................
Control Rod Insertion Limits,...............................
3/4 1-21 3/4.2 POWER DISTRIBUTION LIMITS TABLE 3.2-1 DNB AND REACTOR COOLANT SYSTEM PRESSURE PARAMETERS......
3/4 2-16 3/4.3 INSTRUMENTATION 3/4.3.1 REACTOR TRIP SYSTEM INSTRUMENTATION.........................
3/4 3-1 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
~
TABLE 4.3-1 REACTOR TRIP SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS..........................................
3/4 3-11 3/4.3.2 ENGINEERING SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION...........................................
3/4 3-15 TABLE 3.3-3 ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION.......................................
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 3/4.4 REACTOR CCOLANT SYSTEM 3.4.4.1 REACTOR COOLANT LOOPS AND COOLANT CIRCULATION Startup and Power Operation................................
3/4 4-1 Hot Standby................................................
3/4 4-2 Hot Shutdown...............................................
3/4 4-3 06/01/84 116 Revision A
I SECTION PAGE Cold Shutdown - Loops Filled...............................
3/4 4-5 Cold Shutdown - Loops Not Filled...........................
3/4 4-6 3/4.4.2 SAFETY VALVES Shutdown...................................................
3/4 4-7 Operating..................................................
3/4 4-8 3/4.4.3 PRESSURIZER................................................
3/4 4-9 3/4.4.4 RELIEF VALVES..............................................
3/4 4-10 3.4.4.5 STEAM GENERATORS..........................................
3/4 4-11
. Pressurizer.............................................
3/4 4-35 Overpressure Protection Systems............................
3/4 4-36 j
3/4.5 EMERGENCY CORE COOLING SYSTEMS 3/4.5.1 ACCUMULATORS Cold Leg Injection.........................................
3/4 5-1 I
Upper Head Injection.......................................
3/4 5-3 3/4.5.2 ECCS SUBSYSTEM - Tavg 350*F..............................
3/4 5-5 3/4.5.3 ECCS SUBSYSTEMS - T 5,350*F.............................
3/4 5-9 t
avg I
3/4.5.4 BORON INJECTION TANK (Uni t 1 Only).........................
3/4 5-11 3/4.5.5
. REFUELING WATER STO RAGE TAN K...............................
3/4 5-12 3/.7 PLANT SYSTEMS l
i 3/4.7.1 TUR8INE' 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)
Main Staam Line Isolation Valves...........................
3/4 7-8 l
l Atmospheric Oump Valve.....................................
3/4 7-8a 3/4.7.2 STEAM GENATOR PRESSURE / TEMPERATURE LIMITATION..............
3/4 7-9 l
06/01/84 117 Revision A
r 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 l
i 3/4.7.5 STANOBY NUCLEAR SERVICE WATER POND.........................
3/4 7-12 3/4.9 REFUELING OPERATIONS 3/4.9.1 BORON CONCENTRATION........................................
3/4 9-1 l
3.4.9.2 INSTRUMENTATION............................................
3/4 9-2 3/4. 9. 8 RESIOUAL HEAT REMOVAL AND COOLANT CIRCULATION H i g h Wa t'e r Le v e l......... /..'.........................'......
3/4 9,10 '
Low Water Level............................................
3/4 9-11 l
l l
l l
06/01/84 118 Revision A
F.:
,a a
i, TABLE 2 TECHNICAL SPECIFICATION PAGES AFFECTED The following pages of the Technical Specifications are affected by this review:
T.S. Pages 2-1, 2
TABLE 2.2-1, T.S. Pages 2-5 2-6 2-7 T.S. Pages 3/4 1-1 3/4 1-2 3/4 1-2a proposed 3/4 1-6 3/4 1-7 l
3/4 1-8 t
3/4 1-9 1
3/4 1-10 3/4 1-11 3/4 1-12 3/4 1-13 l-3/4 1-13a) 3/4 1-20a) 3/4 1-21 T.S. Pages 3/4 2-15 16 TABLE 3.3-1, T.S. Pages 3/4 3-2 3-3 3-4 l
3-5 3-6 j
TABLE 3.3-2 T.S. Pages 3/4 3-9 3-10 r
TABLE 3.3-3, T.S. Pages 3/4 3-16 l
3-17 3-18 3-19 3-20 l
3-21 3-22 3-23
(-
r 06/01/84 119 Revision A O
L
r I
I TA8LE 3.3-4, T.S. Pages 3/4 3-25 3-26 3-27 3-28 3-29 TABLE 3.3-5, T.S. Pages 3/4 3-30 3-31 l
3-32 3-33 T.S. Pages 3/4 4-1 l
4-2 4-3 4-4 4-5 l
4-6 4-6(a) proposed 4-7 4-8 4-9 4-10 4-11
[
4'-36 L
T.S. Pages 3/4 5-1,
_~
l 5-2 j-5-2a) proposed 5-2b) proposed 5-3 5-4 5-4a) proposed 5-4b) proposed 5-5 5-6 l
5-9 5-9 l
5-10 5-11 5-12 l
l T.S. Pages 3/4 7-4 7-5(a) proposed l
7-5(c) proposed 7-8 7-8(a) proposed 7-10 f
7-11 7-12 T.S. Pages 3/4 9-1 9-10 9-11 i
l 9-12 06/01/84 120 Revision A i
L.
l
~
APPENDIX A TECHNICAL SPECIFICATIONS SELECTED RELEVANT REGULATIONS 9 Ss.11 Title 10-Energy mined that there are no unresolved (1XI) The processing, fabrication or safety issues relating to the additional refining of special nuclear material or activities that may be authorized pur. the separation of special nuclear mate.
suant to this paragraph that would rial or the separation of special nucle.
constitute good cause for withholding ar material from other substances by a authorization.
prime contractor of the Department (4) Any acdvities undertaken pursu. under a prime contract for:
ant to an authorization granted under
( A) The performance of work for the this paragraph shall be entirely at the Department at a United States govern.
risk of the applicant and, except as to ment. owned or controlled site; matters determined under paragraphs (B) Research in, or development.
(eX2) and (eX3X11), the grant of the manufacture, storage, testing or trans.
authorization shall have no bearing on the issuance of a construction permit pmadon of. Mc wups or m.
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 2 (Secs.101.185. 65 Stat. 936,958, as amended
/
og (42 U.S.C. 2131, 2233: sec.102. Pub. L 91 190. 83 Stat. 833 (42 U.S.C. 4332); sec. 201, as (11) By a prime contractor or subcon.
amended. Pub. L 93-438. 88 Stat.1242. Pub.
tractor of the Commission or the De..
L 94-79, 89 Stat. 413 (42 U.S.C. 3841); see.
partment under a prime contract or 161 as amenced. Pub. L 83-703,65 Stat. 948 subcontract when the Commfulon de.
H 2 U.S.C. = 01n termines that the exemption of the f 21 FR 335. Jan.19,1958. as amended at 23 prime contractor or subcontractor is FR 8712. Sept. 9,1980; 33 FR 2381. Jan. 31.
authorized by law; and that, under the 1YS08. Ao.
4.
terms of the contract or subcontract.
8 ar. 21.
72:39 74 3, FR 26279. July 18.1974: 39 FR 33::02 Sect. there is adequate assurance that the
~
16,1974: 42 FR =887. May S.1977: 43 FR work thereunder can be accomplished 6924. Feb.17.19783 without undue risk to the public health and safety;
$ 50.11 Ezeeptions and esemptions from (2XI) The construction or operation licensing requireinents.
of a production or utilization facility Nothing in this part shall be deemed for the Department at a United States to require a license for:
government. owned or controlled site.
(a) The manufacture, production or including the transportation of the acquisition by the Department of De*
production or utilization facility to or fense of any utilization facility author
- from such site and the performance of ized pursuant to section 91 of the Act, contract servtces 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 ut!II:stion 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, manuisc-1stration facilities of the types subject ture, storage. testing, or transporta.
to licensing pursuant to section 202 of tion of, atomte weapons or components the Energy Reorgani:stion Act of thereof: or the use or operation of 1 I
1974 8 are involved:
production or utilization facility for the Department in a United States
'The Department fact 11 ties identitled in government owned vehicle or vessel:
secuon 202 are Provided. That such activities are con.
(1) Demonstration !Jculd Metal Fast ducted by a prime contractor of the Breeder reactors when operated as part of tr!e poser reneration facilities of an eteetrie ut111ty system. or when operated in any 1973, wnen operated as part of the power other manner for the purpose of demon. generation fact!!ttes of an electric uttilty strattna the suttantitty for commercial ap.
system. or wnen operated in any other sitest:on of sven a reactor, manner for the purpose of dernonstrattns i2) Other cemonstration nuclear reactors. the suitsoGity for commere.a1 application of except tacie in existence on January 19.
such a reactor.
392 06/01/84 121 Revision A
Chapter I-Nucieer Reguietory Commission j 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 contruction or operation of consumers.
a production or utilization facility by a prime contractor or subcontractor of Issuance of such an exemption shall the Commisalon or the Department not be deemed to constitute a commit.
under his prime contract or subcon. ment to issue a construction permit.
s 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 subcontrator is. (b), any activities conducted shall be authorized by law; and that, under the carried out in such a manner as win terms of the contract or subcontract, minimize or reduce their environmen.
there !s adequate assurance that the-tal impact.
work thereunder can be accomplished
[37' FR 574a. Mar. 21.1972. as ame'nded at without undue risk to the public 39 FR 26279. July 18.1974: to FR 8789. Mar.
hesith and safety.
3.1975)
(c) The transportation or possession of any production or utilization faciU. 8 50.13 Attacks and destructive acts 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 FR 8784. Mar. 3.1973) lization facility, or for an amendment to such license. Is not required to pro.
I 58.13 Specifle esemptions.
vide for design features or other mess.
(a) The Commission may, upon ap.'
ures for the specific purpose of protec.
plication by any Interested person or tion against the effects of (a) attacks i
upon its own initiative, grant such ex.
and destructive acts. Including sabo.
emptions from the requirements of tage, directed against the facility by the regulations in this part as it deter.
an enemy of the United States. Wheth.
' mined are authorized by law and win er a foreign government or other j*
not endanger life or property or the person, or (b) use or deployment of common defense and security and~are weapons incident to U.S. defense activ.
otherwtse in the pubtle interest.
itles.
(b) Any person may request an ex*
(32 FR 13445. Sect. 28.19671 emption permitting the conduct of ac.
tivities prior to the issuance of a con
Ltcznsas The Commission may grant such an -
i exemption upon considering and bal.
I 30.20 Two cleases of Ileenees.
ancing the following factors:
Licenses will be Lssued to named per.-
(1) Whether conduct of the proposed sons applying to the Commission activttles win give rise to a significant l
therefor, and w1U be either class 104 or adverse impact on the environment.
and the nature and extent of such class 103.
I I
"**""I*'*
e e redress of any adverse therner and research and development environment impact from conduct of I'* I""' ***
i the proposed activttles can reasonably be effected should such redress be nec.
A class 104 license will be !asued, to essary:
an applicant who qualifies, for any one (3) Whether conduct of the proposed or more of the foUowing: to transf er or activities would foreclose subsequent receive in interstate commerce, manu.
d 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 medical therapy; or by the proposed facility, the availabil.
(b)(1) A production or utili:stion fa.
I ity of alternative sources. If any, to cility the construction or operation of 393 i
i 06/01/84 122 Revision A
+.,
,e (4) The information described in minimum information
- to be included paragraphs (&X1) and (2) of this sec-shall consist of the following:
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 application as provided in is to be located, with appropriate at-paragraph (b) and in accordance with tantion to features affecting facility 12.101 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 after 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 acceptabt!!ty 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 app!!ca' facility will be operated at the ulti-tion for a class 103 construction D*"II*
mate power level which is contemplat.
ed by the applicant. With respect to "O
d)
Y e n who applies for a class 103 construction permit for a nu-power level, the appilcant is required clear power reactor pursuant to the to submit information prescribed in provisions of I 2.101(a-1) and Subpart paragraphs (ax2) through (8) of this P of Part 2 of this chapter shall section, as well as the'information re-submit the document title "Informa-quired by this paragraph, in support tion Requested by the Attorney Gen, of the application for a construction eral for Antitrust Review" at least D'NA L-
_ ~
nine (9) months but not more than (2) A summary descrtption and dis-thirty six months prior to the filing of cussion of the facility, with special at.
part two or part three of the applica. tention to design and operating char-tion, whichever part is filed first, as acteristics, unusual or novel design specified in i 2.101(a-1) of this chap. features, and principal safety consider-
- ter, allons.
(e) Any person who applies for a (3) The preliminary design of the fa-class 103 construction permit for a cility including:
uranium enrichment or fuel reprocess-(1) The principal design criteria for ing plant shall subnut such informa-the faci!!ty.' Appendix A. General tion as may be requested by the Attor. Design Crtterta for Nuclear Power ney General for antitrust review, as a Plants, establishes minimum require-separate document as soon as possible ments for the principal design criteria and in accordance with i 2.101 of this for water. cooled nuclear power plants
- chapter, similar in design and location to plants (Sec.102. Pub. L,91 190. 83 Stat. 353 (42 for which construction permits have U.S.C. 4332n sec. 201 as amended. Pub. t.
previously been issued by the Commis-93 43a. as stat.1242. pud. L 94 79. 39 Stat, alon and provides guidance to app!!-
413 842 U.S.C. 3stin cants for construction permits in es-(29 FR 34398. Sept. 25.1974, as amended at tablishing principal design criteria for 42 FR 22ss7. May S.1977; 42 FR 23721. May other types of nuclear power units; 19.1977: 43 FR 49775. Oct. 23.1978; 44 FR 6071s. Oct. 22.19791
- The applicant may provide information 9 50.34 Contents of applications: technical required by this paragraon in the form of a Information.
discussion, with specific references. of samt.
larltles to and differences from. facilities of (a) Preliminary safety analltsis similar design for wnten applications nave reporf. Each application for a con-previously been filed with the Commission.
struction permit shall include a pre-
- Oeneral design criteria for chemical liminary safety. analysis report. The processms facitattes are belne developed.
399 06/01/84 123 Revision A
$ 50.34 TlHe 10-Energy (ii) The design bases and the rela.
the quality assurance program for a tion of the design bases to the princi.
nuclear power plant or a fuel repro.
pat design criteria; cessing plant shall include a discussion (111) Information relative to materi-of how the applicable requirements of als of construction, seneral arrange. Appendix B will be satisfied.
ment, and approximate dimensions.
(8) An Identification of those strue.
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 analysis and evalu.
tion and description of the research stion of the design ano performance and development program which will of structures, systems, and compo.
be conducted to resolve any safety nents of the faciHty with the objective questions associated with such strue.
of assessing the risit 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.
ment program showing that such tion of (l) the marsins of safety during safety questions will be resolved at or normal operations and transient condi.
before the latest date stated in the ap.
tions anticipated during the life of the plication for completion of construc.
facility, 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 accidenta. Analysis and regulations in this chapter.
evaluation of ECCS cooling 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 sha!! be included in these 150.46 of this part for facilities for plans.
which construction permita may be (11) On or after February S.1979, issued after December 28,1974, applicants who apply for construction (S) 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 hasards to the structures, are determined as the result of pre.
systems and componenta important to
!!minary safety analysis and evalua.
safety of operating nuclear facilities tion to be probable subjecta of techni.
from construction activities. A discus.
cat specifications for the facility, with sion shall also be included of any man.
special attention given to those items aserial and administrative controis which may significantly influence the that will be used during construction final deslan: Provided, hoicever. That to assure the safety of the operating this requirement is not applicable to unit.
an application for a construction (b) Final safety analysis report, permit filed prior to January 16.1969.
Each application for a license to oper.
(6) A preliminary plan for the appil. ste 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.
- 47) A description of the quality as.
tion that describes the facility, pre.
surance program to be applied to the senta 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, componenta 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 planta and fuel re. teorological monitoring programs, processing plants. The description of which has been developed since issu.
400 06/01/94 124 Revision A
Chepeer I-Nuclear Raf-4 Commission i 50.34 ance of the construction permit, relat-(S) A description and evaluadon 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 idenufted 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 requirementa.
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 (1) For nuclear reactors, such items Criteria for Nuclear Power Plants and as the reactor core, reactor coolant Fuel Reprocessine Plants." sets forth system. Instrumentation and control the requirements for such controls for systems, electrical systems, contain-nuclear power planta and fuel repro-ment system. other engineered safety cessing planta. The information on the features, auxiliary and emergency sys-controls to be used for a nuclear power tems, power conversion systems, radio-plant or a fuel reprocessing plant shall active waste handling systems, and include a discussion of how the appit.
fuel hand!!ng systems shall be dis-cable requirements of Appendix B will be us cussed insofar as they are pertinent.
g or preoperational testing (11) For facilities other than nuclear andI" al [ n$ct of normal op-reactors. such items as the chemical, gy3 physical, metallurgical, or nuclear erstions, including maintenance, sur-process to be performed, instrumenta-veillance, and periodic testing of struc-tion and control systems, venuladon tures, systems, and components.
and filter systems, electrical systems.
(v) Plans for coping with emergen-auxiliary and emergency systems, and cies, which shall include the items radioactive waste handling systems specified in Appendix E.
shall be discussed insofar as they are (vi) Proposed technical specifications l perunent.
prepared in accordance with the re-l (3) The kinds and quantitles of ra* quirements of I S0.36.
dioactive materials expected to be pro *
(vill On or after February S.19'f9 duced in the operation and the means applicants who apply for operating 11 for controlling and limiting radiosctive cernes for nucler.r powerplants to be effluents and radiation exposures operated on multtunit sites shall in-
.within the limita set forth in Part 20 clude an evaluation of the potential of this chapter, hasards to the structures, systems, and (4) A finalgglatysis and evaluaHaqof componenta important to safety of op-the design and performance of strue.
ersting units resulting from construc-tures, systems, and components with tion activities, as well as a description the objecttre stated in paragraph of the managerial and administrative f aM4) of this section,gI111Jakins Inte controls to be used to provide assur-acc9MDA.iny. pertinent _mformation de-ance that the limiting conditions for vejoped_since thbubmittal of the pre-operation are not exceeded as a result Ilminary.34Laty_a04)yllt.
Analy.
of construction activities at the mul-sia and evaluation of cooling tiunit sites, performance following postulated loss-(7) The technical qualifications 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 i 50.44 for facilities for resulations in this chapter.
which a license to operate may be (8) A description and plans for im.
Issued after December 28.1974.
piementation of an operator requallfl.
401 06/01/84 125 Revision A
j 50.34a Title 10-Energy connected to the containment atmos-tions thereof, that underlie the corre.
phere. (II.E.4.1) sponding SRP acceptance criteria.
(vil) Provide a description of the (3) The. SRP wasjssusuozs,tablish management plan for design and con-criteria).thaf ths.NRC.ataff intends to struction activitics, to include: ( A) The uss'.lk evaluating'whether-an-appil-organizational and management struc-cant /lleensee me-'
"- Casamission's ture singularly responsible for direc-regulations.The"SRP-ias nos-a mtatt.
Lion of design and construction of the tute for the regulations... anti compil.
proposed plant; (B) technical re-ance is not.a.requiremener AppHeants sourecs director by the applicant; (C) shall identify differences from the details of the Interaction of design and SRP acceptance criteria and evaluate construction within the applicant's or how the proposed alternatives to the ganization and the manner by which SRP criteria provide an acceptable the applicant will ensure close integra-method of complying with the Com.
tion of the architect engineer and the mission's regulatidna.
nuclear steam supply vendor: (D) pro-posed procedures for hand!!ng the (Secs.161b.1811. Pub. L 83-703. 68 Stat.
948: secs. 201, 204(bM1) Pub. L 9 transition to operation; (E) the degree Stat.1242,1243.1245 (42 U.S.C.,..,3-438. s8 01.5841, of top level management oversight and
$844); sec. 7. Pub. L 93-377. 88 Stat. 475:
technical control to be exercised by sec.1611. Pub. L 83 703. 68 Stat. 948 (42 the app!! cant during design and con
- U.S.C. 2201H struction including the preparation (33 FR 18812. Dec.17.1968. as amended at and implementation of procedures 34 FR 6o37. Apr. 3.1969: 34 FR 6770. Apr.
necessagto guide the effort. (II.J.3.1) 23.1969: 38 FR 10499. June 27,1970 33 FR (g) Conformance~tetift ffte Standard 19567. Dec. 24.197o: 36 FR 3256. Feb. 20.
~
lb m. TL ws. M) appucawns 1971: 38 FR 4861. Mar.13.1971: 36 FR fof,1fght " 'a*
~'" """"" Dousac 18201. Sept.11.19711 -
plant operating 6 En:TontAL Notu: For additional recenA;.
t '.~
..m shall Lnclude an evalua. Rectsfun ettations affecting 150.34 see the tion of the facility against the Stand.
1.lst of CFR sections Affected in the Finding ard Review Plan (SRP) In. effect on Aids section of this volume.
May 17,1982 or the SRP revision in 9 50.34a Design objectives for equipment effect six months prior to the docket to control releases of radioactive mate.
date of the application, whichever is fi'l I"
'fII"'""*""""* I P' ***
later.
til) Applications for !!ght water cooled nuclear power plant construe.
(a) An application for a permit to tion permits, manufacturing licenses. construct a nuclear power reactor and pre!!minary or final design appro-shall include a description of the pre.
vals for standard plants docketed after 11minary 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.
effluents produced during normal re-vision in effect six months prior to the actor operations, includ.ng expected docket date of the application, which. operational occurrences. In the case of ever is later.
an application f!!ed on or after Janu.
(2) The evaluation required by this' ary 2.1971. the appi! cation shall also section shall include an identif! cation Identify the design objectives, and the and desertption of all differences in means to be employed. for keeping design features, analytical techniques, levels of radioactive matertal In ei.
and procedural measures proposed for fluents to unrestricted areas as low as a facility and those corresponding fes. is reasonably achievable. The term "as tures. techniques, and measures given low as is reasonably achievable na in the SRP acceptance criteria. Where used in this part means as low as la 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 Improvements 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 1
e j 50.36 Title 10-Energy (b) A construction permit will consti-tions. The technical specifications will tute an authortzstion to the applicant be derived from the analyses *mt aval.
I to proceed with construction but will untion included in the safety _ analysis j
not constitute Commission approval of
- u. ano amenome_ntaJhereta sub-u g
the safety of any design feature or m pursuant to j 50.34. The Com-missfoli' m' Tlhcliiiile such additional specification unless the applicant spe-a I
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:
construction permit or, from time to (17 Sg/ety limif.s. limiting safety time, by amendment of his construc. system settings, and limiting control tion permit. The Commission may, in settings. (1)( A) Safety limits for. nucle-its discretion, incorporate in any con. ar reactors are limits upon important.
struction permit provisions requiring p the applicant to furnish periodic re., rocess variables which are. found to.
be necessary to reasonably protect the ports of the progress and results of re-integrity of certain of the physical search and development programs de.
barriers which guard against the un.
signed to resolve safety questions.
controlled release of radioactivity. If (c) Any construction permit will be any safety !!mit 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 amendment to condition and the basis for corrective the application the complete final action taken to preclude reoccurrence.
safety analysis report, portions of Operation shall not be resumed until 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-
~
mission has found that the final ing plants are those bounds within design provides reasonable assurance which the process variables must be 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 intes.
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.188. 68 Stat. 955: 42 U.S.C. 22351 trolled release of radioactivity. If any (27 m 12913. Dec. 29.1992, as amended at safety limit for a fuel reprocessing 31 FR 12780. Sept. 3o,1966; 35 m 5318, plant is exceeded, corrective action Mar. 31.1970: 35 FR 6644. Apr. 23.1970 33 shall be taken as stated in the technt.
FR 11441. July 7.19701 cal specification or the affected part of the process, or the entire process if I 50.36 Technical specifications, 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 !!censee sha!!
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 than those covering administrative entire process has been shut down. op-controls, shall also be inrluded in the erstion shall not be resumed until au.
application, but shall not beenme part thorized by the Commission.
of the technical specifications.
(!!HA) Limiting safety system set.
(b) Each license authortzing oper.
tings for nuclear reactors are settings ation of a production or utilization fa.
for automatic protective devices reist.
ellity of a type described in 150.21 or ed to those variables having signif!.
I 50.22 will include technical specifica.
cant safety functions. Where a limit.
410 06/01/84 127 Revision A
[ +,*.
Chapter 1-Necleer R:C y 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 case of either a chosen that f
- 'c protective nuclear reactor or a fuel reprocessins action will correct the abnormal situs-plant, the licensee shall notify the tion before a safety 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-ciuding the esuse of the condition and quired, the licensee shall take appro-the basis for corrective action taken to 4
priate action, which may include sjtut-preclude reoccurrence.
tingdown the reactor. He shall notify (3) Surveillance requirements. Sur-tiie 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 of systems and components is taken to preclude reoccurrence.'
maintained. that facility operation will (B) Limiting control settings for fuel be within the safety limits, and that reprocessing plants are settings for the limiting conditions of operation automatic alarm or protective devices will be met.
related to those variables having sig.
(4) Design features. Design features i
nificant safety functions. Where a to be included are those features of Ilmiting control setting is specified for the facility such as materials of con.
a variable on which a safety limit has struction and geometric arrangements.
been placed, the setting shall be so which. If altered or modified, would 3
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. during operation.
d
( )
ifstrative controls. Admin-v ces or, unc n u
the Istrative controls are the provisions re.
licensee shall take appropriate action i
to maintain the variables within the lating to organization and manage.
limiting control. setting values and to ment, procedures, recordkeeping.
repair promptly the automatic devices review and audit, and reporting neces-or to shut down the affected part of sary to assure operation of the facility the process and, if required. to shut in a safe manner.
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 Commission, review the fleations included in any license issued matter, and record the results of the prior 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 conditions for oper.
report as technical spectfications.
l siton. Limiting conditions for oper.
(2) An spplicant for a license author.
ation are the lowest functional c.3ng 12ing operation of a production or uti.
bytty.,,gr performance TFvels of equi (.
lization facility to whom a construe.
ment required for safe ope'rEon og tion permit has been issued prior to theTscility. When s limiting condittori January 16, 1969, may submit techni-for operation of a nuclear reactor is can specifications in accordance with not met, the !!censee shall shut down this section, or in accordance with the i
the reactor or follow any remedial requirements of this part in effect action permitted by the technical speca prior to January 16.1969.
(3) At the initiative of the Commis.
ification until the condition can be met. When a limiting condition for op.
sion or the licensee, any license may erstion of any process step in the be amended to include technical spect.
system of a fuel reprocessing plant is i fications of the scope and content not met, the !!censee shall shut down which would be required if a new 11 that part of the operation or follow cense were being issued.
I 411 1
4 Il l
)
06/01/84 128 Revision A i
J
_.,___v__
E.
s Chapter I-Nuclear Reguletery Commission
{ 50.42 4 $0.3A Incilrihility of certain applicanta.
(d) Any applicable requirements of Any person who is a citizen nation-Part 51 have been satisfied.
al. or agent of a foreign country, or [21 FR 355. Jan.19.1956, as amended at 36 any corporation or other entity which FR 12731. July 7.1971: 39 FR 26279. July the Commission knows or has reason 18.1974: 47 FR 13754. Mar. 31,19621 to believe is owned, controlled. or.
dominated by an allen, a foreign cor-8 50.41 Additional standards for class 104 poration. or a foreign government,
- licenses, shall be ineligible to apply for and In determining that a class 104 li-obtain a license.
cense will be issued to an applicant.
(Sec.161. as amended. Pub. 1.83-703. 68 the Commission will. In addition to ap-Stat. 948 (42 U.S.C. 2201); sec. 201, as plying the standards set forth in amended. Pub. 1.93-438. 88 Stat.1243 (42 6 50.40 be guided by the following con-U.S.C. 5841n siderat!ons:
[21 FR 355. Jan.16.1956 as amended at 43 (a) The Commission will permit the FR 6924 Feb.17.19781 widest amount of effective medical therapy possible with the amount of 9 50.39 Public inspection of applications, 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 mr.de availa-conduct of widespread and diverse re-ble for public inspection in accordance search and development.
with the provisions of the regulations (c) An application for a class 104 op-contained in Part 2 of this chapter.
ersting !! cense as to which a person who intervened or sought' by timely STANDARDS roR LterNsEs AND written notice to the Commission to CoNstaucTtoM Pram Ts intervene in the construction permit 550.10 Common standards.
proceeding for the facility to obtain a
~-
deter 7nination of antitrust consider-In determining that a license will be attons or to advance a jurisdictional issued to an app!! cant, the Commis* basis for such determination has re-sion 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 FEDanAt. RrotsTrn of notice of filing of and equipment, the use of the facility.
the application for an operating 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 provt.
foregoing collectively provide reason
- sions of I 50.42(b).
able assurance that the applicant will 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 safet of the pub!!c will not be endan.
9 $0.12 ' Additional standards for eiaas 103 (b) The applicant is technically and II""*"-
financially quallfled to engage in the In ' determining whether a class 103 proposed activities in ac ordance with license will be issued to an appilcant.
the regulations in this chapter. How. the Commission will, in addition to ap.
ever, no consideration of financial plying the standards set forth in qualifications is necessary for an elec. I 50.40, be guided by the following tric utility applicant for a !! cense for a considerations:
production or utillt.ation facility of the (a) The proposed activities will serve type described in 150.21(b) or 150 2'!.
a useful purpose proportionate to the (c) The issuance of a Itcense to the quantitles of special nuclear material applicant will not, in the opinion of or source material to be utilized.
the Commission, be lnimical to the (b) Due account will be taken of the common defense and security or to the advice provided by the Attorney Gen.
health and safety of the public, eral, pursuant to subsection 10$c of 413 06/01/84 129 Revision A m
s s,
Chapter I-Mueleee Rzn y Commission 950.4 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 be repressurized beyond 50 per. from containment, through filters if
- cent of the containment design pres. necessary, such as purgmg systems and repressunzation systems and (11) sure.
(g) For facilities with respect to systems that do not result in a signiff.
which the ' otice of hearing on the ap.
cant release from containment such as n
recombiners.
plication for a construction permit was 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.
ment atmosphere to the erntronment the low population zone outer bound.
through filters if needed.
ary from purging (and repressuriza.
(4) A repressunzation 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 contam.
lated in accordance with 1100.11(aN2) ment by adding inert gas or air to the of this chapter is less than 25 rem to containment. Dilution of the combus.
the whole body and less than 300 rem tible gas results in a delay in time to the thyroid, only a purging system untH a flanmtable conemtration is is necessary, provided that the purging reached and permits fission product system and any filtration system asso. decay. Operation is limited to a con.
cisted with it are designed to conform tainment npmsunzation to 50 per.
with the general requirements of Crt. cut of t e containment design pm.
terta 41,42, and 43 of Appendix A to mm. A purgmg system is nonnaHy l
this part. Otherwise, the facility shall part of e npmse.zation system.
be provided with another type of com-bustible gas control system (a repres. (Sec.161 as amended. Pub. L 83 703, 68 sunzation system is acceptable) de.
Stat. 948 (42 U.S.C. 2201): sec. 201, as amended. Pun. L 93-438. 88 Stat.1*42. Pub.
signed to conform with the general re.
- t. 94-19,89 Stat. 413 (42 U.S.C. S4411) qutrements of Cnteria 41,42, and 43 of Appendix A to this part. If a purge [43 rit 30163. Oct. 27.1978, as amended at 44 FR 38444. Dec. 2,1981]
system is used as part of the repressur.
ization system. It shall be designed to conform with the general require. 8 50.45 Standards for construction per.
ments of Cnteria 41,42, and 43 of Ap.
mits.
pendix A to this part. The contain.
An applicant for a license or an nient shall not be repressurtzed amendment of a license who proposes 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 entena of means that the performance of the 11 50.31 through 50.38 and the stand.
emergency core cooling system ta pos. ards of ll 50.40 through 30.43.
tulated for purposes of design of the combustible gas control system, not to 1 30.46 Acceptance criteria for emergency meet the acceptance ertterta in 150.46 core cooling erstems for light water and that there could be localized clad mlear Po*H rese**.
melting and metal. water reaction to (aN1) Except as provided in para.
the extent postulated in paragraph (d) graph taH2) and (3) of this section.
of this section. The degree of perform-each bo!!!ng and pressurtzed 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/S4 130 Revision A
4 a
e o
{ 50 4 Title 14 3%
lindrical Zircaloy cladding shall be complete it. The Director of Reguta-provided with an emergency core cool.
tion of the Atomic Energy Commission ins system (ECCS) which shall be de.
shall have caused notice of such a re.
signed such that its calculated cooling quest to be published promptly in the performance following postulated loss.
Fansaat RasssTan; such notice shan of-coolant accidents conforms to the have provided for the suomission of criteria set forth in paragraph (b) of comments by interested persons this section. ECC5 cooling perform.
wrthin a time pertod established oy ance shall be calcuisted in accordance the Director of Regulation. If ucon with an acceptable evaluation modet, reviewing the foregoing submissions and shall be calculated for a number the Director of Regulation concludeo of postulated loss of-coolant accidents that good cause. had been shown for of different sizes locations, and other an extension. he may have extenceo properties sufficient. to provide assur.
the six month period for the shortest ance that the entire spectrum of pos.
additional time which in his judgment tulated loss of coolant accidents is cov.
will be necessary to enable the licensee ered. Appendix K. ECCS Evaluation to furnish the submissions required by Models, sets forth certain required and paragraph (aH2Xiu of this section. Re.
acceptable features of evaluation questa for extensions of the six month models. Conformance with the criteria period submitted under this subpart.
set forth in paragraph (b) of this sec.
graph will have been ruled upon by tion with ECCS cooling performance the Director of Regulation prior to ex.
calculated in accordance with an ac.
piration of that period.
ceptable evaluation model, may re.
(iv) Upon submission of the evalua-quire that restrictions be imposed on tion required by parsgrsph (aH2HID of reactor operation.
this section (or uncer parsgraoh (2) With respect to reactors for (aM2X11D. if the six month pertad '.s which operating licenses have prevt. extended) the facility shall contim:e ously been issued and for which oper. or commence operation only withm ating licenses may issue on or before the limits of both the proposed technt.
December 28.1974:
cal specifications or !! cense amend.
(D The time within which actions re. menta submitted in accordance with quired or permitted under this part. this parsgraph (aH2) and all technical graph (aH2) must occur shall begin to specifications or license conditions run on February 4.1974.
preytously imposed by the Atomte (11) Within six months following the Energy Commission. Including the re.
date spectf!ed in paragraph (aH2XD of quirements of the Interim Po11cy this section an evaluation in accord. Statement (June 29 1971. 36 FR ance with parsgraph (aHI) 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 that the evaluations submitted under proposed changes in technical spectfl. parsgraphs (aH2) (11) and (lid of this estions or license amendments as may section are not consistent with pats-be necessary to bring reactor oper. graph (aX1) of this sect!on and as a atton in conformity with paragraph result such restrictions are required to
(&H1) of this section.
protect the public health and safety.
f111) Any licensee may have request.
(vD Exemptions from the operating ed an extension of the six month requirements of parsgraph (aH2Hivi period referred to in paragraph of this section may be granted for f aH2H!D of this section for good cause. good cause. Requests for such exemp.
Any such request shall have been sub. tion shall be submitted not less than mitted not less than 43 days prtor.. 45 days prior to the date upon which expiration of the six month period, the plant would otherwise be required and shall have been accompanied by to operste in accordance with the pro.
affidavits showing precisely why the cedures of said paragisch (&H Hly) of evaluation is not complete and the this section. Any such request shall be mtntmum time believed necessary to filed with the Secretary of the Com.
418 06/01/84 131 Revisien A i
Chapter 8-Nwleer Reguletery Commission I 50.44 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 ox1 the FEDrRAL RzotsTra; 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 FrDrnAL Rro. tance from inside to outside the clad-IsTra publication. The Director of Nu.
ding, after any calculated rupture or clear Reactor Regulation shall submit swelling has occurred but before sig.
his views as to any requested exemp. n'ficant. oxidation. Where the calculat-tion within five days following expira-ed conditions of transient pressure and tion of the comment pertod.
temperature lead to a prediction of (v11) Any request for an exemption cladding swelling, with or without submitted under paragraph. (ax )(vi) cladding rupture, the unoxidized clad-of this section must show, with appro. ding thickness shall be defined as the priate affidaytts 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 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 average circum.
(a)( Hivl of this section. The request ference at that elevation. For ruptured shall also include a discussion of the cladding the circumference does not alternatives available for establishing include the rupture opening, compliance with the rule.
(3) Martmum Avdrogen 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 before December 28, 1974 subject tion of the cladding with water or to any app!! cable 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 Core Coollag Systems published on fuel, excluding the cladding surround.
June 23,1971 (36 FR 12:48) as amend-ing the plenum volume, were to react.
ed (December 18.1971. 36 FR 24082):
(4) Coolable geometry. Calculated Pfortded. Actceter 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 (S) Long ter'n cooling. After any cal-this paragraph. unless the Commission culated successful initial operation of determines, among other things that the ECC3 the calculated core tem-the proposed facility ratets the re. perature shall be maintained at an ac.
quirements of paragraph (aH1) of this ceptably low value and decay heat section.
shall be removed for the extended (bH1) Peak cladding temperature.
perted of time required by the long-The calculated maximum fuel element lived radioactivity remaining in the cladding temperature shall not exceed core.
- 00* F.
(c) As used in this section: (1) Loss.
(2).tfa:tmum cladding o.itdation, of coolant accidents (LOCNs) are hy The calculated total oxidation of the pothetical accidents that would result cladding shall nowhere exceed 0.17 from the loss of reactor coolant. at a times the total cladding thickness rate in excess of the capact11ty of the before oxidation. As used in this sub-reactor coolant makeup system. from paragraph total oxidation means the breats in pipes in the reactor coolant total thickness of cladding metal that pressure boundary up to and including 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 stoichiometrac altecnfum (2) An evaluation model is the calcu.
dioxide. If caldms rupture is calculat. lational framewora for evaluating the 419 06/01/81 132 hvision A
s
{ 50.g TlHe le '
.g, behavior of the reactor system during finding will constitute a rebuttable a postulated loss of coolant accident presumption on questions of adequacy (LOCA). It includes one or more com-and implementation capability. Emer-
'i puter programs and all other informa-gency preparedness exercises (required tion necessary for application of the by paragraph (b)(14) of this section calculational framework to a specific and Apper:diz E. Section F of this LOCA. such as mathemmHe*I models part) are part of the operational in-used, assumptions included in the pro-spection process and are not required grams, procedura for treating the pro-for any initiallicensing decision.
gram input and output information.
(b) The onsite and. except as pro-specification of those portions of anal.
vided in paragraph (d) of this section, ysis 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, the 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 been specifically established, and each implementation of the general re-principal response organization has quirements with respect to ECC3 cool-staff to respond and to augment.its ing performance design set forth in initial response on a continuous basis.
this part. Including in particular Crite-rion 35 of Appendix A.
(2) On shift facility licensee respon-sibilities for emergency response are (39 FR 1002 Jan. 4.1974. as amended at 39 unambiguously
- defined, adequate FR 271:1. July 23.1974: to FR 8789. Mar. 3.
stalfing to provide initial facility acci-18731 dent response in key functional areas
-d M is maintained at all times, timely sus-IM7
- s. __
mentation of 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.
NRC that there is reasonable assur-(3) Arrangements for requesting and ance that adequate protective mens-effectively using assistance resources ures can and will be taken in the event have been made, arrangements to ac-of a radiological emergency.
commodate State and local staff at the (2) The NRC will base its finding on licensee's near site Emergency Oper-a review of the Federni Emergency ations Facility have been made, and Management Agency (FEMA) findings other organizations capable of aug-and determinations as to whether menting the planned response have State and local emergency plans are been identified.
adequate and whether there is reason-(4) A standard emergency classifica-able assurance that they can be imple-tion and action level scheme, the bases mented, and on the NRC assessment of which include facility system and as to whether the applicant s onsite effluent parameters, is in use by the emergency plans are adequate and nuclear facility licensee, and State and whether there is reasonable assurance local response plans call for reliance that they can be implemented. A FEMA finding will primartly be based on a review of the plans. Any ot! frin-iThese standards are addressed by specif-Ic enteria in NUREO-o654: FEMA-REP-1 formation already available to FEMA enuded Cntena for Preparation and Eval-may be considered in assessing wheth-uation or Raatoiosical Emersener Response er there is reasonable assursnee that Plans and Preparedness in support of Nucle-the plans can be implemented. In any ar Power Plants-for Intenm Use and Com-NRC licenstng proceeding, a FEMA ment". January 19so.
420 06/01/84 133 Revision A
Chapter I-Nucleer Reguletery Commission g 50.37 authortsed by the Comminion upon request Footnotes to i 50.55a:
pursuant to i 50.55stax:)(u).
' [ Reserved!
' For purocess of this regulation. the pro-posed IEEE 2"9 became "in effect" on Components which are connectM to the 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.
reactor coolant pressure boundary def!nd Copies may be obtained from the Institute in iso.2(v) need not meet these require-of Electrical and Electronics Enstneers.
ments, provided:
United Enstneering Center. 345 East 47th (a) In the event of postulated faalure of Street. New Yort. NY 10017. A copy is avaal.
the cosapenent.durma normal reactor oper-able for inspection at the Comminaion's ation, the Macta can be shut down and Pubue Document Room.1717 E Street NW cooled down in an orderly manner. Samming Washington. D.C.
makeup is provided by the reactor coolant
- Where an application for a construction up system only. w permit is submitted in four parts pursuant (b) The component is or can be isolated to the provtsions of l 2.101(a-D and Subpart from the reactor coolant system by two F of Part : of this chapter. "the formal valves (both closed. both open, or one ciceed doctet date of the application for a con-and the other open). Each open valve must struction permat" for purposes of this sec.
be capable of automauc actuation and, as.
tion shall be the date of docteting of the in-suming the other valve is open its cloom formauon required by 12.101(a-1)(2) or.(3),
time must be such that. In the event of pos-whicheve is later.
tulated failure of the component during normal reactor operation each valve r,.
I 50.54 Conversion of construction permit mains operable and the reactor can be shut to IIcense: or amendment of license.
down and cooled down in an ordwty or alterat,mpletion of the construction Upon co manner, assuming makeup is provided by
.on 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 Amd.
can Society of Mec1= *al Enstneers, the construction permit and subject to United Engineertas Center. 345 East 47th EDy necessary testing of the facility St New York. NY 10017. Copies are availa-for health or safety purposes, the ble for inspection at the Nawaa's Commies!On W1U. in the absence of Pubuc Document Room.1717 H St. NW good cause shown to the contrary Waahtnarton. D.C.
!ssue a UCense of the Class for wh!Ch
'USAS and A8ME Code addenda issued the construction permit was issued or prior to the Winter 1977'Addanda are con-an appropriate amendment of the U-sidered to be "tn effect" or " effective" 6 Cense. as the case m&Y M.
months after their date of tasuance and after they are incorporsted by reference in (Sec.188. 64 Stat. 955; 42 U.S.C. :35) paragraph (b) of this section. Addanda to the ASME Code tasued after the Summer (21 FR 338. Jan.19.1954, as amended at 33 1977 daada are considered to be "In FR 11441. July 17.19701 effect" or " effective" after the date of pubu-5 50.57 Issuance of operating license.'
cauon of the addenda and after they are in-(a) Pursuant to i 50.56. an operating corporated by reference in parasraph (b) of this section.
Ucense may be issued by the Commis-
'For ASME Code Editions and Addenda sion, up to the full term authorized by tasued prior to the Winter 1977 Addenda. ! 50.51 upon finding that:
the Code Edition and Addenda appucable to (1) Construction of the faciuty ha.s the component is soverned by the order or been substantiaUy completed, in con-contract date for the component. not the contract date for the nuclear enersy system. formity with the construction per=.1t and the application as amended, me For the Winter 1977 addends and subse-quent editions and addenda the method for provisions of the Act. and the rules determining the appucable Code editions and regulations of the Commission:
and addenda is contained in Paragraph NCA and 1140 of Section III of the ASME Code.
8 The Commisston may issue a provtsional
'ASME Code cases which have been de-oceaung ucense pursuant to the Ms.
termined suitable for use by the Commis-cons 2 tMs part in edut on ha 30.
sion staff are listed in NRC Regulatory 1970. for any fact!!ty for which a notice of Guide 1.84. " Code Case Acceptability-hetrms on an application for a provtstonal ASME Section III Destsn and Fabr! cation-opersung ucense w a notice of preposed >
and NRC Regulatory Guide 1.35. " Code suance of a provtstonal operating license has Case Acceptability-ASME Section !!I Ms.
b"a pubushed on w tefwe that date.
tenais." The use of other Code cases may be 437 06/01/84 134 Revision A
. - - - ~. --
a e
4 f50J8-Title 10-Energy (2) The facility will operate in con-this section as to wnich there is a con-formity with the application as troversy, in the form of an initial deci-amended the provisions of the Act.
sion with respect to the contested ac-and the rules and regulations of the tivity sought to be authorized. The Di-6 Commission; and rector of Nuclear Reactor Regulation (3) There is reasonabic assurance (!)
will make findings on all other matters that the activities authorized by the specified in paragraph (a) of this sec-operating license can be conducted tion. If no party opposes the motion, without endangering the health and the presiding officer will issue an safety of the public, and (11) that such order pursuant to 12.730(e) of this activities will be conducted in compil-chapter, authorizing the Director of ance with the regulations in this chsp-Nuclear Reactor' Regulation t9 make ter;and appropriate findings on the matters (4) The app!! cant is t,echnically and specified in paragraph (a) of this sec-financially qualified to engage in the; tion and to issue a license for the re-activities authorized by'the operating quested operation.
license in accordance with the regula-tions in this chapter. However, no (35 FR 5318. Mar. 31.1970. as amended at finding of financial qualifications is 35 FR 6644. Apr. 25,1970; 37 FR 11873. June 15.1972: 37 FR 15142. July 28.1972: 40 FR necessary for an electric utility appi!*
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 150.21(b) or I 50.22.
g 50.58 Ilearings and report of the Advine-(5) The app!! cable provisions of Part ry Committee on Reactor Safeguards.
140 of this chapter have been satisfied apphh for a hm (6) The issuance of the license will tion permit or an operating license for not be inimical to the common defense a facility which is of a type described and security or to the health and in l 50.21(bl or l 50.22. or for a testing safety of the public.
facility, shall be referred to the Advi-(bs Each ope. sting license wiu In.
sory Committee on Reactor Safe.
clude appropriate provisions with re, guards for a review and report., An ap-spect to any unconspleted items of plication for an amendment to such a construction and such limitations or construction permit or operating 11-conditions as are required to assure cense may be referred to the Advisory that operation during the period of Committee on Reactor Safeguards for the completion of such items will not review and report. Any report shall be I-endanger public health and safety, made part of the record of the app!!ca-(c) An applicant may. In a case tion and available to the public. except 4
where a hearing is held in connection to the extent that security classifica-with a pending proceeding under this tion 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 ersting license authorizing low power and publication once in the FrntnAL testing (operation at not more than 1 Rzorsrsa on each application for a percent of full power for the purpose construction permit for a production of testing the facility), and further op.
cr utilization facility which is of a erstions short of full power operation, type described in 150.21(b) or 150.22 Action on s ach 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 I
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 appiles-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 the activity to be authorized. Prior to permit or operating !! cense, the Com-taking any act!on 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 finrnas on the once in the FrntaAL Rzorstra or, in matters specified in paragraph (a) of the absence of a request therefor by 438 06/01/84 135 Revision A
+-
a y
y--
,--2--
--g-a7-
r
--w
o a
y Chapter 1-Nuclear llegulatory Commission j 50.70 any person whose interest may be af.
ments carried out pursuant to para-fected, may issue an operating license graph (a) of this section. These ree-or an amendment to a construction ords shall include a written safety pennit or operating license without a evaluation which provides the bases hearing, upon 30 days notice and pub-for the determination that the change.
!! cation once in the PenenAL Racistra test or experiment does not involve an of its intent to do so. If the Commis-unreviewed safety question. The 11-sion finds that no significant hazards censee shall furnish to the appropriate consideration is presented by an appil-NRC Regional Office shown in Appen-cation for an amendment to a con-dix D of Part 20 of this chapter with a struction permit or operating license, copy to the Director of Inspection and it may dispense with such notice and Enforcement. U.S. Nuclear Regulatory publication and may issue the amend-Commission. Washington, D.C. 20555 t
- ment, annually or at such shorter intervals
. ~ '
(27 FR 12186. Dec. 8.1962. as ame' ded at 33 as may be specified in the license, a n
FR 8590. June 12,1968: 35 FR 11461. July report containing a brief description 17.1970: 39 FR 10555. Mar. 21.19741 of such changes, tests, and experi-ments. Including a sumraary of the 3 50.39 Chanees. tents and experiments.
safety evaluation of each. Any report (a)(1) The holder of a license submitted by a licensee pursuant to authorizing operation of a production this paragraph will be made a part of or utilization facility may (1) make the public record of the licensing pro-changes in the facility as described in ceeding. In addition to a signed origi-the safety analysis report. (!!) make nal. 39 copies of each report of changes in the procedures as described changes in a faci!!ty of the type de-in the safety analysis report, and (!!!)
scribed in 150.21(b) or i 50.22 or a 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, without prior Commission approval.
shall be filed. The records of changes unless the proposed change, test or ex*
In the facility shall be maintained periment involves a change in the until the date of termination of the 11-technical specifications incorporated cense, and records of changes in proce-In the license or an unreviewed safety dures and records of tests and experi-question.
ments shall be maintained for a period (2) A proposed chanse, test, or ex* I of five years.
periment shall be deemed to involve (c) The holder of a lic.ense author-an unreviewed safety question (!) If the probability of occurrence or the Izing operation of a production or utt-
!!sation facility who desires (1) a consequences af 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 the procedures described in the safety safety analysis report may be in.
creased; or (11) if a possibility for an analysis report or to conduct tests or accident or malfunction of a d!!ferent experiments not described in the type than any evaluated previously in sdiety analysis report, whic.h involve a
the safety analysis report may be cre-an unreviewed safety question or a ated; or (!!!) 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 license pursuant to j 50.90.
(b) The licensee shall maintain ree-(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 3.1976; 4: FR 20139. Apr.18.19771 to this section. to the extent that such changes constitute changes in the fa.
Isserc rtons. Rzconos. F =onts, cility as described in the safety analy.
Nortricartons r
sia report or constitute changes in pro-9 50.70 Inspections.
cedures as described in the safety analysis report. The licensee shall also (a) Each licensee and each holder of maintain records of tests and experi-a construction permit shall permit in-439 06/01/84 136 Revision A
r gys8 4
j 50.90 Title 10-Energy with the regulations in this chapter REVOCATION. SUsrENstoN. MoorricA.
and will not be inimical to the TION. AMENDMENT or LICENSES AND common defense and security or to the CONSTRUCTION PERutTs. EutacENCY health and safety of the public.
OPERAT!oNs BY THE Commission (b) If the application demonstrates that the dismantling of the facility $ 50.100 Revocation. suspension. modifica-l and disposal of the component parts tion of licenses and construction per-I will be performed in accordance with mits for cause.
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 fense and security or to the health whole or in part, for any material false and safety of the public. and after statement in the application for 11 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 dismant!!ng 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 to refuse to grant a license on an origi-
[ S FR 9546. Oct.10.1961 as amended at 32 nal application (other than those re-FR 3090. Fee. 21.19671 lating to Ii 50.51, 50.42(a),
and AMENDMENT or LICENst oR CoNsT!!UC-50.43(b) of this part): or for fa !ure to Tr6N PEnMIT AT request or Hot. Den construct or operate a facility in ac-cordance with the terms of the con-4 50.20.typtication for amendment of li.
struction permit or license, provided cense or construction permit.
that failure to make timely completion of the proposed construction or alter-
.Whenever a holder of a !! cense or stion of a facility under a construction construction permit desires to amend permit shall be governed by the provi-the license or permit, application for stans of 150.55(b): or for violation of.
an amendment shall be filed with the or failure to observe, any of the terms Commission. fully describing _~the and provisions of the act regulations, changes desired. and following as far
!! cense, permit, or order of the Corn-a1 appficable the form presenbed for mission.
original applications; i
1 50.101 Retaking possession of special 9 50.51 Issuance of amendment.
nuclear material In determining whether an amend.
Upon revocation of a license. the ment to a license or construction Commission may immediately cause permit will be issued to the applicant the retaking of possession of all spe-the Commission w(!! be guided by the cial nuclear material held by the !!.
considerations which govern the issu; censee.
ance of Initial licenses or construction permits to the extent app!! cable and (21 FR 355. Jan.19,1956, as amended at 40 FR 8790. Mar. 3.19751 appropriate. If the application in-volves the material alteration of a 11 4 30.102 Commission order for operation censed facility, a construction permit after revocation.
will be issued prior to the issuance of the amendment to the license. If thel 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 reduction program of the Depart-pursuant to i 2.105 of th chapter ment requires continued operation of before acting thereon. The notice will a production or util!2ation facility, the be issued as soon as practicable after license for which has been revoked.
the application has been docketed.
the Commission may, after consulta-tion with the appropriate federal or (39 FR 13:58. Apr.12.19741 state regulatory agency having juris-444 l
06/01/84 137 Revision A
e Chapter I-Nuclear Regulatery Comminion Port 50, App. A diction, order that possession be taken compliance with the rules. regulations, of such facility 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 !! cense to submit such in-duction program of the Departrient formation concerning the addition or may require. or until a license for op-proposed addition, the elimination or eration of the facility shall become ei-proposed elimination. or the modifica-fecitve. Just compensation shall be tion or proposed modification of strue.
paid for the use of the facility.
tures, systems or components of a fa-(40 FR 8790. Mar. 3.19751 cility as it deems appropriate.
S 50.103 Suspension and operation in war or national emergency.
ENroRCEMENT (a) Whenever Congress declares that a state of war or national emergency 9 50.110 Violations.
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.
tion of any provision of the Atomic (1) Suspend any license 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 !!-
Issued 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 recapture any rule, regulation, or order issued of special nuclear material cr by oper-thereunder, cr any terrn, condition, or ation of any facility, pursuant to this !!mitation of any !! cense issued there-sectlen.
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) 186 of the Act. Any person who will-(21 FR 355. Jan.19.1956, as amenced at 35 fully violates any provision of the Act PR 11416. July 17,1970: 40 FR 8790. Mar. 3.
or any regulation or order issued 19751 thereunder may be guilty of a crime and, upon convictim. may be punished BAcmTtwo by fine or imprisonment or both, as 5 $0.109 Ilackfitting.
provided by law.
(a) The Commission may. in accord-I40 FR 8790. Mar. 3.1975, as amended at 42 ance with the procedures specified in Fi} 257:1. May 19.19771 this chapter, require the backfitting of
$rrtNmcEs a facility if it finds that such action will provide substantial, additional protection which la required for the public health and safety or the ArrgNorx A-GENEttAt. DESIGN common defense and security. As used CaITERIA ron Nuct. EAR PowEn PLANTS in this section. "backfitting" of a pro-M **M8""""
duction or utilization facility means the addition, elimination or modifica-twTnonectson tion of structures, systems or compo-nents of the fac!!!ty after the con.
otrmmons struct!on permit has been issued.
.Nue! car Power Unit.
(b) Nothing in this section shall be Loss of Coolant Accidents.
deemed to relieve a holder of a con-Single Failure.
struction permit or a license from Anticipated Operational Occurrences.
445 06/01/84 12G Revision A