ML20066B328
| ML20066B328 | |
| Person / Time | |
|---|---|
| Site: | McGuire, Mcguire  |
| Issue date: | 01/02/1991 |
| From: | Licciardo R Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20066B322 | List: |
| References | |
| TAC-55435, TAC-55436, TAC-67757, NUDOCS 9101070069 | |
| Download: ML20066B328 (101) | |
Text
.-.
4 DIFFERING PRDFESSIONAL OPINION (DPO)
REVIEW 0F MCGUIRE TECHNICAL SPECIFICATIONS DATED JUNE 11, 1984 BY ROBERT B.A. LICCIARDO SAFETY EVALVATION COMMENTS ON REVIEW BY THE NRC ENTITLED
" CLOSURE OF DP0 ISSUES REGARDING MCGUIRE TECHNICAL SPECIFICATIONS (TACS 55435/55436/67757)"
DATED SEPTEMBER 10, 1990 PREPARED BY ROBERT B.A. LICCIARDO PLANNING, PROGRAM AND MANAGEMENT SUPPORT BRANCH OFFICE OF NUCLEAR REACTOR REGULATION JANUARY 02, 1991 Robert B.A. Licciardo Registered Professional Engineer, California Nuclear Engineering License No. NV 1056 Mechanical Engineering License No, M 015380
\\
D DO K O O
69 PDH P
i
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EXECUTIVE
SUMMARY
PART 1 OVERALL REVIEW The detailed and overall evaluation undertaken by the writer during his safety evaluation of the Dr. Murley memorandum of September 10, 1990, shows that the original categorization by the Reactor Systems Branch in respect of 0;,en Items, is no longer valid and that the writer's original Safety Evaluation of the Proof and Review of the MCGUIRE TS's as represented in the REVIEW OF MCGUIRE TECHNICAL SPECIFICATIONS, dated June 11, 1984 was and remains valid for all items and that the potential number of concerns which may be closed by later clarification without any additional licensing action is evaluated at only about 6%.
The original number of items identified as concerns were totalled by the NRC as
- 380, out of which it selected 220 items for review for incorporation into either Plant Specific and/or Generic Ts, and thereby excluding 160 which are identified in the table by the symbol (0) under the column "0)EN". In the final i
analyses, a total of 421 items were identified out of which 174 (0) items were identified (instead of 160)
From 421 items 308 were ultimately evaluated out of which a 264 necessary licensing actions were identified, The remaining 86 residual items are valid and.
should now be considered; the writer's safety evaluation of these items remain unchanged PART 2 PLANT SPECIFIC ISSVES l
Fifty one items of concern by the writer in his DP0 Review were identified as Plant Specific Issues, and of these 48 items have or will require plant-specific or ' generic action in the form of amendments to the TS, FSAR, IST, and SPM for McGuire, and including 15 items for inclusion in the NSTS of which 5 should be added to the WSTS. Three (3) items only were closed out completely by licensee _ clarification a_ lone representing only 6% of the total Plant _
specific concerns and which thereby establishes the validity of his McGuire TS review to Ref. A.1 in respect of these safety concerns.
In conclusion the level' and quality of the NRC review has not been that expected from a Peer group review of the writer's safety concerns for the McGuire Facility at -a point in time which is nine years after the commencement of operations of the Facility PART 3 GENERIC ISSUES The total number of items identified for generic consideration by a minimal set l
of various entities is 240.
The total number of necessary additions to the NSTS is 207 and of this count many are also included in the WSTS under the same item numbers (CINS) i
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e The total number of CINS impacted by both changes to the ETS and or the WSTS is 170, and would represent the total impact on the Existing Tech. Specs, alone for the MCGUIRE UNITS and which would have protected the plant against the Mid-Loop loss of Residual Heat Removal Cooling at both the Diablo Canyon and Votgle Units In conclusion the level and quality of the review in the area of Generic concerns is Unacceptable as a Peer group review for the writer's McGuire TS Review of Ref A.1 l
1 i
ii
e 5
f TABLE OF CONTENTS EXECUTIVE
SUMMARY
LIST OF TABLES TABLE NOMENCLATURE PART 1:
OVERALL COMMENTS ON THE NRC REVIEW 0F THE R.B.A. LICCIARDO MCGUIRE TS REVIEW OF 1984 (REF. A.1)
PART 1.1:
SUMMARY
PART 1.2: DISCUSSION OF TABLE 1 TABLE I PART 2:
COMMENTS ON THE PLAN 1 SPECIFIC REVIEWS OF THE R.B.A. LICCIARDO MCGUIRE TS REVIEW 0F 1984 (REF. A.1)
PART 2.1
SUMMARY
PART 2.2 DISCUSSION OF TABLE 2 PAP.T 2.3 DETAILED COMMENTS ON THE PROPOSED CLOSE OUT OF THE PLANT SPECIFIC ACTIONS BY THE T.M. MURLEY MEMO OF SEPTEMBER 10 1990.
1 TABLE 2 PART 3:
COMMENTS ON THE REVIEW OF GENERIC DP0 ISSUES OF THE R.B.A. LICCIARDO MCGUIRE TS REVIEW 0F 1984 (REF. A.1)
PART 3.1
SUMMARY
PART 3.2 DISCUSSION ON MINIMAL SET OF GENERIC ACTIONS FOR THE NSTS ARISING FROM VARIOUS ENTITIES: REF. TABLE 3.1 PART 3.3 DISCUSSION ON MINIMAL SET OF ACTIONS ON THE EXISTING AND WESTINGHOUSE I
TS ARISING FROM VARIOUS ENTITIES, AND DERIVING FROM THE R.B.A.
l LICCIARDO MCGUIRE TS REVIEW OF 1984 (REF. A.1):
REF. TABLE 3.2 PART 3.4 DETAILED COMMENTS ON THE REVIEW OF THE PROPOSED CLOSE OUT OF GENERIC ISSUES BY THE T.M. MURLEY MEMO 0F SEPTEMBER 10 1990 TABLE 3.1 TABLE 3.2 LIST OF REFERENCES v
e s
l LIS1 0F TABLES TABLE 1:
TOTAL LIST OF CONCERNS FROM THE R.B.A. LICCIARDO MCGUIRE TS REVIEW 0F 1984 (REF. A.1) LISTED BY (CONGRESSIONAL) ITEM NO (CIN):
RECORD OF REVIEWS BY DIFFERENT ENTITIES TABLE 2:
CINS EVALUATED AS PLANT SPECIFIC BY A. THADANI MEMO OF MAY 14,1990 AND RELATED PLANT SPECIFIC (PS) (AND GENERIC) RESOLUTIONS (A SUBSET OF TABLE 1)
'ABLE 3.1:
LIST OF MINIMAL SET OF GENERIC ACTIVITIES AND ACTIONS ARISING FROM VARIOUS ENTITIES, AND DERIVING FROM THE R.B.A. LICCIARDO MCGUIRE TS REVIEW 0F 1984 (REF. A.1). (A SUBSET OF TABLE 1)
TABLE 3.2:
LIST OF MINIMAL SET OF ACTIONS ON THE EXISTING TS AND WESTINGHOUSE STS ARISING FROM VARIOUS ENTITIES, AND DERIVING FROM THE R.B. A.
LICCI ARDO MCGUIRE TS REVIEW OF 1984 (REF. A.1).
(A SUBSET OF TABLE 1),
vi i
4 O
1 l
LIST OF TABLES TABLE NOMENCLATURE GENERAL:
4 1)
INDICATES TITLES TO COLUMNS 2)
OTHER THAN TITLES: REMAINING NOMENCLATURE IDENTIFIES THE CONGRESSIONAL ITEM NUMBr.RS (CINS) FOR WHICH THERE ARE i
RELATED ACTIONS TOGETHER WITH ADDITIONAL INFORMATION ON THEIR ORIGIN DETAIL:
A ASHOK THADANI ACTION OF JUNE 1990 MEMO ACXX CODIFICATION OF-GROUPS OF ITEMS FROM THE ORIGINAL REACTOR SYSTEMS BRANCH SELECTION OF 220, BY THE DIVISION OF LICENSING IN ITS REVIEW TO REFERENCE A.3 A
CIN LOCATOR FOR PLANT SPECIFIC ELEMENTS OF THE ORIGINAL REVI." 0F 220 ITEMS WHICH REMAINED OPEN FOR RESPONSE BY THE LICENSEE.
EITHER GENERIC OR PLANT SPECIFIC ISSUES. ACTIUN ARISING FROM THADANI LETTER OF MAY 14 1990
.A CIN LOCATOR FOR PLANT SPECIFIC ELEMENTS OF THE ORIGINAL REVIEW OF 220 ITEMS WHICH REMAINED'0 PEN FOR RESPONSE BY THE LICENSEE, AND-SUBSEQUENTLY SELECTED FOR CONSIDERATION AS GENERIC ISSVES FOR INCORPORATION INTO THE NEW STANDARD TECHNICAL SPECIFICATIONS AND OR THE WESTINGHOUSE STANDARD TS ACTION ARISING FROM THADANI LETTER OF MAY 05 1990
..A CIN LOCATOR-FOR ELEMENTS Of THE ORIGINAL REVIEW OF 220 ITEMS WHICH REMAINED OPEN FOR GENERIC CONSIDERATION FOR INCLUSION IN THE WESTINGHOUSE STANDARD TS OR THE NEW STANDARD TS. ACTION c
ARISING FROM THADANI LETTER OF MAY 14 1990 TO REF. 37 "ASHTAD" ACTIONS FROM ASHOK THADANI LETTER OF MAY 14,1990 (REF. 37)
BCXX CODIFICATION OF GROUPS OF ITEMS, FROM THE ORIGINAL 220 ITEMS-SELECTED BY THE-FORMER REACTOR SYSTEMS BRANCH, BY THE DIVISION OF LICENSING IN ITS REVIEW TO REFERENCE A.3 L
CSA CIN LOCATOR FOR ITEMS CLOSED SATISFACTORILY AND LATER ADDED TO l
TS ("CLOSSATADD")
CC CINS CLOSED BY LATER CLARIFICATION BY LICENSEE, WITH OR WITHOUT ADDITIONS TO TS,FSAR,SPM OR IST.
"CLOSUS" CINS CLOSED UNSATISFACTORILY BY DL'S REVIEW.T0 REF. A.3-- RBAL'S EVALVATION.
CI-CIN LOCATOR FOR "CLOSINVALI" "CLOSINVALI" CLOSVRE INVALID--RBAL'S ORIGINAL EVALVATION OF DL REVIEW 0F REF. A.3.
"CLOSLATARRE" CLOSED BY DL REVIEW 0F REF.A.3.
FOR LATER REVIEW BY RBAL.
"CLOSSATISF" DELETED "CLOSSATADD" CLOSED SATISFACTORILY AND LATER ADDED TO THE MCGUIRE TS, FSAR, SPM, OR IST, L
l vii l
e 6
"CLOSUNSOP" CLOSED UNSATISFACTORILY, BUT PHILOSOPHICAL DISCUSSIONS IN PROGRESS BY DL REVIEW OF REF. A 3 "eLOSCLAR" CLOSED WITH CLARIFICATION.
[
CLR CIN LOCATOR FOR CLOSLATRRE CV CIN LOCATOR FOR "CLOSUNSOP" l
EQRE CINS FOR WHICH RELATED EQUIPMENT HAS BEEN REMOVED i
ETS CINS FOR EXISTING PLANT SPECIFIC TS - AMENDMENT AGREED TO ETS+
CINS FOR EXISTING PLANT SPECIFIC TS - AMENDMENT IS NECESSARY
)
ETSW CINS FOR ADDITION TO ETS AND WSTS - AMENDMENT IS AGREED TO L
ETSV+
CINS FOR ADDITION TO ETS AND WSTS - AMENDMENT IS NECESSARY f
"EXISTTS" EXISTING TS WHETHER HCGUIRE OR WSTS "FSAR" FINAL SAFETY ANALYSIS REPORT FSA CINS FOR FINAL SAFETY ANALYSES REPORT-AMENDMENT AGREED TO FSA+
CINS FOR FINAL SAFETY ANALYSES REPORT-AMENDHENT IS NECESSARY G
CIN LOCATORS FOR GENERIC ISSUE G.RSW CIN LOCATORS FOR GENERIC ACTION RSB G.RSB CIN LOCATORS FOR GENERIC ISSUE BY REACTCR SYSTEMS BRANCH L
" GENERIC" ITEMS IDENTIFIED AS GENERIC BY VARIOUS ENTITIES "GENSTUDY" GENERIC STUDY: GENERIC LETTERS, OWNER'S GROUPS. W STUDY, NRC STUDY p
"GENERICWE" GENERIC TO WESTINGHOUSE TS.
-GSWN CINS FOR GENERIC STUDY BY WESTINGHOUSE AND NRC i
GW CINS FOR WESTINGHOUSE GENERIC ISSUE IST CIN FOR IN SERVICE INSPECTION PROGRAM-AMENDMENT AGREED TO I
" ITEM N0".
CONGRESSIONAL ITEM N0 4
i "LRH45" TS REQUIREMENTS AND CONCERNS FOR LOSS OF RESIDUAL HEAT REMOVAL IN l
MODES 4 AND 5.
LRS CINS FOR LRH45 "LRH6" TS REQUIREMENTS AND CONCERNS FOR LOSS OF RESIDUAL HEAT REMOVAL IN 1
MODE 6.
LR6 CINS FOR LRH6 "HEWSTS" NEW STANDARD TS-SELECTED FOR CONSIDERATION NSTS CINS FOR WHICH REVIEWERS HAVE AGREED TO.
NSTS+
CINS FOR WHICH NSTS ADDITION IS NECESSARY "0 PEN" IDENTIFIES 160-ITEMS EXCLUDE 0 FROM REVIEW BY BERNER0 MEMO OF AUGUST 30, 1984, REF. A.16 O
CINS FOR "0 PEN" ITEMS EXCLUDED BY REF A.16 "0THACTN":
-ACTION BY OTHER ENTITIES "PLNT SPEC".
SELECTED AS A PLANT SPECIFIC ITEMS of THE DIVISION OF LICENSING IN ITS REVIEW TO REFERENCE A.3 PS CIN FOR PLANT SPECIFIC ITEM RSBWSTS CIN FOR SUBMITTAL BY REACTOR SYSTEMS BRANCH FOR INCLUSION IN WESTINGHOUSE STANDARD TS's i
viii
~.
l "RSBSELECN" RSBS SELECTION: CODIFICATION OF GROUPS OF ITEMS FROM THE CRIGINAL REAC10R SYSTEMS BRANCH SELECTION OF 220, BY THE DIVISION OF LICENSING IN ITS REVIEW TO REFERENCE A.3 "SETONTHETH" SET POINT METHODOLOGY SPM CIN FOR SET POINT METHODOLOGY-AMENDMENT AGREED TO SPM+
CIN FOR SET POINT METHODOLOGY-AMENDMENT IS NECESSARY "TMS10" THOMAS MURLEY AUTHORIZATIONS ARISING FROM REFERENCES 37 AND 40 TMF CINS TOR TMS10 ITEMS FROM REF. 40 TM CINS F0st TM510 ITEMS FROM REF. 37 W
CINS FOR: INDIVlvuAl. ACTIONS BY WESTINGHOUSE.
WL CINS FOR: WESTINGHOUSE LETTER RECOMMENDATIONS TO ALL WESTINGHOUSE REACTOR OWNERS V0G LOCATES CINS FOR STUDY BY WESTINGHOUSE OWNERS GROUP WSTS CINS FOR ITEMS INSIDE WESTINGHOUSE STANDARD TS W.WSTS CiNS FOR WESTINGHOUSE RECOMMENDA1 ION FOR WSTS iX
j t
PART 1:
OVERALL COMMENTS ON THE NRC REVIEW OF THE R.B.A. LICCIARDO MCGUIRE TS REVIEW OF 1984 (REF. A.1)
PART 1.1
SUMMARY
The original number of items identified as concerns were totalled by the NRC as 380, out of which it selected 220 items for review for incorporation into either Plant Specific and/or Generic Ts, and thereby excluding 160 which are identified in the table by the symbol (0) under the column "0 PEN". Ir the final analyses, a total of 421 items were identified out of which 174 (0) items were identified (instead of 160)
From 42. M ms, 308 were ultimately evaluated out of which a 264 necessary licensing actions were identified. The remaining 86 residual items are valid and should now be considered; the writer's safety evaluation of these items remain unchanged PART 1.2. DISCUSSION OF TABLE 1--TOTAL LIST OF CONCERNS FROM THE R.B. A. LICCIARDO MCGUIRE TS REHEli 0F 1984 (REF. A.1) LISTED BY (CONGRESSIONAL) ITEM NO (CIN):
RECORD Of *EVIEWS BY DIFFERENT ENTITIES This table lists the totai number of Items of Concern raised by the writer in his original DP0 and the totality of related NRC reviev activities.
It also identifies a minimal set of generic actions undertaken by various entities since the preparation of the McGuire-TS Review.
These includes Generic
.etter 86-13 by the NRC To All Power Reactor Licensees And Applicants With Combustion Engineering and Babcock and Wilcox Pressurized Water Reactors, A destinghouse letter to All Westinghouse Licensees on the subject of the Number Of Reactor Coolant Pumps In Mode 3.
A Westinghouse Owners Group Action on LOCA in Mode 4; items schmitted for inclusion in the Westinghouse Standard TS by the former Reactor dystems Branch; particular Generic studies by W and the NRC, Particular TS submittals by W through Licensees on overpressure protection in Modes 3 through S, and other items as identified in the Table Nomenclature.
f The o_riginal number of items identified as concerns were totalled by the NRC as 380, out of which it selected 220 items for review for incorporation into either Plant Specific and /or Generic Ts, and thereby excluding 160 which are identified in the table by. the symbol (0) under the column "0 PEN". In the final L
analyses, a total of 421' items were identified out of which 174 (0) items were
(-
identified (instead of 160)
A detailed evaluation of the table will show that in the overall analyses to l
date, 88 of these 174 Open Items were identified for evaluation as primarily generic concerns leaving a residual of 86:
So that out of 421 items, 308 were ultimately evaluated out of which a detailed analysis of the table will show that'264 necessary licensing actions were identified. Furthermore, a detailed check of the-86 residual items by the groups into which they remain show the items are valid and should now be considered; the writer's safety evaluation of these items remain unchanged Of the total No of 421 items, only six items were closed by clarification and by principally the licensee.
The remainder remained valid.
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TABLE 1 (cont) l 1
1
~. - - -. -
-o PART 2:
COMMENTS ON THE PLANT SPECIFIC REVIEWS OF THE R.B.A. LICCIARD0 MCGUIRE TS REVIEW 0F 1984 (REF.1)
PART 2.1
SUMMARY
Fifty one items of conc:rn by the writer in his DP0 Review were identified as Plant Specific Issues, and of these 48 items have or will require plant specific or generic action in the form of amendments to the TS, FSAR, IST, and SPM for McGuire, and including 15 items for inclusion :in the NSTS of which 5 should be added to the WSTS.
Three (3) items only were closed out completely by licensee clarification alone representing only 6% of the total Plant e
specific concerns and which thereby establishes the validity of his McGuire TS review to Ref. A.I. in respect of these safety concerns.
In conclusion the level and quality of the NRC review has not been that expected from a peer group review of the writer's safety concerns for the McGuire Facility at a point in time which is nine years after the commencement of operations of the Facility L
PART 2.2 DISCUSSION OF TABLE 2: CINS EVALUATED AS PLANT SPECIFIC BY A. THADANI MEMO 0F MAY 05,1990 AND RELATED PLANT SPECIFIC (PS) (AND GENERIC)
RESOLUTIONS (A SUBSET OF TABLE 1)
A-detailed analysis of Table 1 will show that of the 220 items selected for review, 127 items were originally identified as plant specific and these are recognized in Table 1 by the identifier PS under the Column "PLNTSPEC." in the final analysis.
By the A. Thadani memo. of May 14, 1990, this number was further reduced to 51 by re-identification-of the remaining items for generic consideration.
The residual PS items and the consequences of these are shown in Table 2.
The Writer's evaluation of the most recent Murley Memo to Ref. 40 is included.
Arising from the evaluation of Generic Issues identified by the Reviewers, the table showsithat 48 items have or will require plant specific or generic action
.in the form of amendments to the TS, FSAR, IST, and SPM for McGuire, and includ-ing 15 items for inclusion in the NSTS of which 5 should be added to the WSTS, l
Three (3) items only were closed out completely by licensee clarification alone.
l L
Of the original 127 PS items, detailed analyses of the table will show that only 7 remain to be reviewed and these remain valid.
l-l l;-
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2-1
PART 2.3 DETAILED COMMENTS ON THE PROPOSED CLOSE OUT OF THE PLANT SPECIFIC ACTIONS BY-THE T.M. MURLEY MEMO 0F SEPTEMBER 10 1990. is a copy of the DEVIEW OF MCGUIRE TECHNICAL SPECIFICATION to Ref.-A.I. This identifies the Congressional Item Numbers (CINS) for each of the items of corcerns reported to the US Congress by the Chairman Nunzio J Palladino of the USNRC by letter of December 20, 1984. is the copy of the T. Murley memorandum to Ref.- 40 and which is'a principal. subject of these Comments.
These detailed Comments are made directly against Enclosure 3, sub-enclosures 1, 2, and 3, and particularly against each of the items-in the sequence in which they are presented in that document.
Where no further comment is made the item is not generally addressed: Reference to this document is essential.
For Plant Specific concerns the NRC review to Ref. 40 has suffered from a number of deficiencies, nd these are detailed in the Comments. Summarily, they include the following:
In considering a set of disparate events with a particular common safety characteristic, Reviewers have oversimplified the evaluation by focussing on only one event and in a manner from which generalized invalid conclusions are derived for the remaining events In considering single events, the Reviewers have considered a less than minimum partial set of the information required for the safety analyses of the events and have thereby faulted in their safety evaluation and-in specifying a less than minimal set of such information for inclusion in the TS's Some Reviewers have revealed a singular lack of the necessary detailed knowledge of related Regulatory requirements for all Protective systems including manual operations thereto, for the reactor, and the facility in general. This has lead to speculation on important features of events and the necessary protective responses by operating staff and including the availability of protective equipment and the preparation of procedures including the TS.
The Reviewers conti_nually pr'opose positions outside the licensing bases for the McGuire units and which are thereby invalid. They have also l
misread the Writer's evaluation.
They have also made evaluations based l
on faulted knowledge of the Protective logic.
The licensee has made L
faulted statements which have taken three cycles of review to be finally accepted by the licensee for ultimate correction.
Questions directed to resolving specific issues have been ignored causing p
them to remain open There is a marked lack of capability in the necessary detailed Nuclear Engineering of the facility and especially the Protective systems, to meet Regulatory Requirements 2-2 1!.-.
=.
s And likewise there is a marked and thereby very serious lack of capability in the understanding, and thereby establishing of, the detailed safe operation of the_ reactor under normal operating conditions, to ensure licensing basis-safety under T&A conditions.
There-is a particular unwillingness by the HRC staff to enforce _ Regulatory Requirements.for the evaluation of changes to Set Point Methodology in the H
form of Safety Analyses Limits and related Margins and in a manner which leaves _TS for Set Points and Allowable Values unchanged.
The fctmal eval-uation by Amendment of TS changes required by the regulations is established by the fundamental protection policy that all actions associated with the determining the safe operation of the plant through the TS are so important' I
that the-NRC must be detail checked for Acceptance irrespective of its effect on the margin to safety.
By proposing not to do this for many of' the items in the the sub-enclosure 3 of Enclosure 3 is a non-conformance of _NRC Regulatory-_ responsibility.
Furthermore all values-_of parameters important to the safe operation of the Plant as determined by safety analyses:are required to be reported in the FSAR; and this has not been enforced by Reviewers.
In conclusion the level and quality of the NRC review has not been that expected from a Peer group review.of the writer's safety concerns for the McGuire facility at a point in time which is nine years after the commencement of operetions of the Facility 2-3
. >0 S t Detailed Comments by'R.B.A. Licciardo on T. Murley Closure of OPO Issues Regarding the Mc Guire Technical Specifications, dated September!10, 1990.
Question 15, TS 3/4.5.3-
RESPONSE
The licensee has accepted the writer's proposal to have only centrifugal charging pump operable below the specified temperature and together with the provision that the safety injection pumps are rendered incapable of delivering to the RCS.
This proposal is acceptable provided that_both of these constraints are included in-the-TS.
In setting related TS Temperature limits below which this is necessary, licensee should have a set point methodology which recognizes not only the errors in the Tavg measuring system, but also the differences in temper-ature between the location in Reactor Coolant Pressure Boundary-where the critical stress / temperature limits will occur, and the Tavg being mea-sured and used as indicator of that temperature. _In this respect it ap-pears that the general-limit of 350 deg. F which is the value-used to categorize TS Section 3/4.5.3 in the Standard TS has been chosen with those considerations in mind and which also answers the writer's original concern as to the TS provision for more pumps being available between 350-deg. F and 300 deg. F, than at lesser temperatures.
So that effectively the temperature at which only one pump only, and that is the centrifugal Pump, should be capable of injecting into the RCS should be 350 deg. F. as read on the Tavg measuring system.
2-4 Robert B.A. Licciardo December 1990
.,=...
o Detailed Comments by R.B. A. Licciardo on T. Murley Closure of_OPO Issues _Regarding the Mc Guire Technical Specifications, dated September 10, 1990.
QUESTION 4.c. Table 3.3.2, Item 17:
REACTOR TRIP INSTRUMENTATION RESPONSE TIMES.
RESPONSE
The~" clarifications" undertaken by Reviewers in the course of the Mc Guire TS Review have clouded a number of important fundamental issues which re-main unresolved.
And these are clearly stated in the writer's TS Review and substantively elaborated upon with license basis information in his memo of June 10, 1990, Ref. 14.
Except for the very-limited clarification of terminology, and related arnendment to the TS, in the reviewer's responses the rest of the resolution to this particular question is unacceptable and for very significant areas of necessary protection over a ccmplete range of break sizes in both the primary coolant system and the Main Steam System.
This is also a generic issue for TS.
The substance of the writer's positions in these areas are fully document-ed in previous submittals, but are repeated here in part because of their importance.
RESOLUTION RBAL_ Position - Reference response under Issue 2 below.
Reference also comments under Questions 7b and 7.
9 Issue 1.
No Response from Licensee The functional Unit described as Safety Injection Input from ESF" is in-correct.
TS descriptors should be replaced by four functional units con-l sistent with Table 3.3-5; i.e., by Manual Safety-Injection, Containment l
Pressure-High, Pressurizer Pressure-Low (SI) and Steam Line Pressure-Low.
Proposed Ae ion: -TS descriptors should be replaced.four functional units consisten' sith Table ~3.3-5;.i.e., by Manual Safety Injection, Containment Pressure-High, Pressurizer Pressure-Low (SI) and Steam Line Pressure-Low.
L Issue No 2.
Related_ Response Times omitted from TS by proposing as Not l
Applicable (N.A).
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The Licensee responds that trip _ functions not utilized in FSAR transient I
and a;cident analyses will have the requirement indicated as Not Applica-
-ble (N.A.).
RBAL Response--This position is incorrect and thereby Unacceptable.
An essential regulatory requirement is diversity of Protection Systems so that all-licensing basis transients and' accidents will in general have at least two separate parameters initiating protective action._ Also Tran-sient & Accident (T&A) analysis will also generally be undertaken with l
-the second out trip, or other later trip, giving the most conservative evaluation considered necessary for the expected consequences of the'Oc-currence.
In this regard it should be noted that for the parameters in 2-5 Robert B. A. Licciardo December 1990
~
Detailed Comments by R.B. A. Licciardo on T. Murley closure of DP0 Issues.Regarding the Mc Guire Technical Specifications, dated September 10, 1990, question, examples include LOCA and MSLB Breaks inside~ and.outside con-tainment, both small and large; and such breaks in modes 3 and 4: For transients, the excessive cool down resulting from failure open of.the main feedwater valves is an event where this is use as back up parameter.
As a first out, or diverse protection, this reactor trip is especially important for events below the P-7 permissive when direct reactor trip from another parameter may not be available.
Proposed Action:
The term NA alongside item 17 in this Table 3.3-2 should be replaced by the response times used in the Accident Analyses.
Note the actual response times are included in Table 3.3-5 and under the more accu-rate descriptors required of Issue 1 above.
Issue No 3.
Absence of docketed information for times used in related Accident Analyses,'and particularly for MSLB, SBLOCA and LOCA events.
Proposed Action: The writer has discovered docketed information and which-
.is different from that of existing TS values.
Reference response to Ques-tions 7b an 7g.
The corrected values should be inserted in this Table 3.3-2,-Item 17.
Issue 4:
This issue has been resolved.
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-i 2-6 Robert B. A. Licciardo December 1990
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- Detailed Comments. tot R.B. A. Licciardo on T. Murley Closure of DP0 Issues Regarding-the Mc Guire Technical Specifications, dated September 10, 1990.
Question Ib, Table 2.2-1, Item 4.
RESPONSE
-The response has ignored the fact outlined in earlier submittals by the writer that-the negative flux rate trip setpoint was not evaluated as part-of the safety analyses for Mc Guire as there was_ no approved Evaluation Methodology for the related Transient.
The setpoint methodology document was indeed in error.
As a result of the DPO,.later_ NRC approved Evaluation Methodology has now been used and the licensee has revised the Setpoint Methodology Table 3-4, to_ show a safety analyses limit of 6.9% rated thermal power.
This value permits the TS trip setpoint:and allowable values to remain unchanged.
Consequently the conclusions should show Amendment to the FSAR to record these changes in the related safety evaluation, and also to the_ related Set Point (SP) Methodology.
This also became a generic issue for Westinghouse units.
2-7 Robert B.A. Licciardo December 1990
Detailed Comments b/ R.B.A. Licciardo on T. Murley Closure of OP0 Issues Regardiig the Mc Guire Technical Specifications, dated September 10, 1990.
Question ic, TS Table 2.2-1.
Item 9
RESPONSE
The response concerning the set point methodology document is invalid.
This document is the only primary source of information on the safety ana-lyses limits on Section 15 Transient & Accident evaluations and as such performs a primary reference in evaluating licensing basis amendments to a 10 CFR 50.59(a)(1) requirements.
And furthermore is the only source of information for checking the Set Points and Allowable Values of the TS.
All changes to the SPM should therefore be by a formal Amendment.
Any current practice which ignores this is irregular and non conforming to regulatory requirements l
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l 2-8 Robert B. A. Licciardo December 1990 l
l
o Detailed Comments by R.B.A. Licciardo on T. Murley closure of DP0 Issues Regarding the Mc Guire Technical Specifications, dated September 10, 1990.
Question Id, TS Table 2.2-1, Item 13
RESPONSE
A number of changes have been made to this particular TS since the writ-er's review.
The original question was valid.
The writer's previous work has revealed that the setpoint specified in the setpoint methodology document was a non-conservative application of the allowance for channel error and drift.
As advised under earlier review, the licensee has changed the bounding analysis event for this parameter to that of the Main Feedwater Line Rup-ture initiating at full power and assuming a low-low water level Safety analyses Limit of 23% of narrow range span.
The licensee now states that the Mc Guire TS setpoint for the SG low-low water level trip, at 100% rat-ed thermal power, "is now 40% of narrow range span which exceeds the safety analyses limit value of 23% narrow range span by more than 10%".
This change in Safety Analysis Limit for the SG should be be reflected in a necessary amendaient to the Set Point Methodology Report for Mc Guire Units 1&2, Ref. 18, and also as a change to the FSAR (from the original value of > or = 54.9%).
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l 2-9 Robert B. A. Licciardo December 1990 l
1
Detailed Comments by R.B. A. Licciardo on T. Murley Closure of OP0 Issues Regarding the Mc Guire Technical Specifications, dated September 10, 1990.
Question le, Table 2.2-1, Item 18b i
RESPONSE
The last descriptor for this Question le, i.e., as " Item 18b", is incor-rect and should be replaced by "18 c(i) (last para)".
The reviewers have not addressed the detailed clarifications by the writer Nevertheless, in his previous submittal and therefor remains incomplete.
the licensee's original response itself remains satisf actory, and no changes are required.
2-10 Robert B.A. Licciardo December 1990
Detailed Comments by R.B. A. Licciardo on T. Murley Closure of DP0 Issues Regarding the.Mc Guire Technical Specifications, dated September 10, 1990.
Question 2, TS Page 3/4 1-6,-(TS 3.1.1.4)-
RESPONSE-
~
The-licensee should be advised that the Qualitative Evaluation provided is Unacceptable in meeting the Regulatory requirements for safety analyses during the proposed experiments under 10CFR50.59, and the arguments based on probability of being within that temperature range is an infringement of TS requirements under 100FR50.36.
The licensee provides a qualitative evaluation which proposes to show that for a MSLB, at End Of Core Life, with negative moderator temperature coef-ficient, nuclear power is reduced when the minimum temperature fo'r criti-cality:is reduced from 557 deg. F to 551 deg. F.
The writer agrees with this proposition.
However, DNBR is ultimately established from a combi--
nation of Thermal-Hydraulic as well as nuclear power condition's; and for the MSLB-the reduction of average temperature from 557~deg. F to 551 deg. F also causes a significant reduction in the reactor vessel pressure _under the resulting thermal hydraulic environment with emptied pressurizer and voiding with flashing in the Reactor Vessel head, so that resulting DNBR -
is reduced even though the return to nuclear power is reduced Ref. 42, Sections 3.3.3.13 and 3.1.4. (Item) Set.
The writer notices that this concern is not restricted only to Mc Guire
~TS's, but also is applicable to all other facilities using Standard TS's, and thereby is a generic issue, i
2-11 Robert B.A. Licciardo December 1990
Detailed Comments by R.B. A. Licciardo on T. Murley Closure of OP0 Issues Regarding the Mc Guire Technical Specifications, dated September 10, 1990.
Question 3, TS Table 3.3.1. Item 6c
RESPONSE
The reviewers have chosen not to respond in a specific and valid manner to the writer's concerns from the plant specific licensing and regulatory requirements for Mc Guire 1 & 2 and have referred to a later generalized letter without plant specific licensing action authority.
The licensee is therefore in violation of his. licensing bases commitments in respect of this item.
The proposed TS's were invalid and remain invalid until they conform to FSAR commitments by having at least two Source Range Neutron Flux channels being operable in Modes 5-3 with effective alarms whilst the reactor trip breakers are in the open position.
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l 2-12 Robert B.A. Licciardo December 1990
Detailed Comments by R.B.A. Licciardo on T. Hurley Closure of OP0 Issues Regarding the Mc Guire Technical Specifications, dated September 10, 1990.
Question Sa, Table 3.3-3, Item 79:
RESPONSE
The licensee has agreed with the proposition that the blockage of the trip in Mode 3 below Mode 3# is not acceptable.
Further, the licensee has ac-cepted the need for operability (of automatic initiation of Auxiliary Feed Water) in Mode 4.
This item on these two subjects is now closed with two (2) changes to the TS.
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2-13 Robert 8,A. Licciardo December 1990 1
Detailed Comments by R.B. A. Licciardo on T. Murley Closure of DP0 Issues Regarding the Mc Guire Technical Specifications, dated September 10, 1990.
Question 6b, Table 3.3-4, Items 7c(1) and (2)
RESPONSE
The necessary clarification of an apparent inconsistency between the TS and the Accident Evaluation is acceptable, and the FSAR should be modified to clarify this issue, The response to the question of flow distribution under accident condi-tions is incomplete as the engineering features of the plant show that some form of flow control device must be used under these circunistances and this information is not provided in the FSAR.
The licensee must pro-vide these details for evaluation and the FSAR.
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l 2-14 Robert B. A. Licciardo December 1990
Detailed Comments by R.B.A. Licciardo on T. Murley Closure of DP0 Issues Regarding the Mc Guiro Technical Specifications, dated September 10, 1990.
Question 6c, Table 3.3.-4, Item 9.
RESPONSE
The reviewers have responded to the question of the basis for the set point concerned, but have not responded to the consequences of that in terms of the residual issue.
The licensee response confirms that the setpoint for the Emergency Busses allows them to be unloaded of all Non-ESF loads during 100% normal opera-tion of the plant, without the reactor being tripped by the Undervoltage Trip on the RCP Busses, and consequently that af ter the Emergency Bus is transferred to DG supply, all of the Non-ESF loads will not be restored and this non-restoration of Non-ESF loads could mean the loss of services necessary for the continuing safe normal operation of the plant.
Since there is no analyses of the consequences of the loss of these services in the Mc Guire FSAR, this represents an unanalyzed condition for the operat-ing reactors at Mc Guire.
The writer is advised that this this is also potentially a generic issue.
This concern should be evaluated and incorporated into the FSAR.
2-15 Robert B.A. Licciardo December 1990 l
Detailed Comments by R.B.A.- Licciardo on T. Murley Closure of DP0 Issues Regarding the Mc Guire-Technical C.pecifications, dated September 10,~ 1990.
Question 7a and 7f; Table ~3.3-5, Item 2a; Table 3.3 5, Item 3a.
RESPONSE
1.
The reviewers have not responded to the fact that LOCA's below P-11 Interlock were evaluated and are a part of the Licensing Bases for Mc Guire Units 1 and 2.-
Reference Question 8e of TABLE 4 of this review concerning rey Item TS 3/4.4.1.
G 2.6.3.
2.
The reviewers have not responded to the items 2, 3, 4 & 5 of the writer's comments of Ref. A.14, except to effectively admit that the existing response time of the TS for the RHR/LOCA pumps is indeed incorrect and to provide an unacceptable justification for that.
The reviewers have also responded by providing response times from the TS, whereas it is the Safety Analyses that provide the bases for these values and no reviewer.has responded to that fact. The current response provides l-no bases for an acceptable resolution.
The licensee shall respond specifically to the details of the writer's earlier review Ref. June 19 of information extracted from his own FSAR, and provide amendments to his TS in accordance with the data provided un-less he has later documentary data to support different values.
Additionally the licensee has furthermore revised his LOCA analyses and leaves related update of the FSAR until 1991.
Since his analyses has al-ready been completed, the NRC should clarify-the core reload status to which it applies, and if applicable to the current condition, immediately require the submission of the appropriate TS Amendments, i
2-16 Robert B.A. Licciardo December 1990
J_W.
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____a a _
Detailed Comments by R.B.A. Lieciardo on T. Murley Closure of OP0 Issues Regarding the Mc Guire Technical Specifications, dated September 10, 1990.
Question 7b and 7 ; Table 3.3-5, Item 2b; Table 3.3-5, item 3b 9
The reviewers have ignored the detailed comments by the writer on this particular issue and therefore I must presume they cannot be answered and therefore remain valid.
This affects a large number of significant TS's and should be closed out with direct responses by the licensee to each of the review's questions without evasions.
The licensee response to these ocerns is incomplete and unacceptable.
Action:
1.
For TS Table 3.? 5, Items 2b, 3b,and 4b, the current descriptor Reactor Trip (f rom SI), must be replaced by only " Reactor Trip".
2.
For TS Table 3.3-5, Items 2b, 3b,and 4b, the current response times of 2 secs. must be replaced by > or = 0.46 secs.
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2-17 Robert B.A. Licciardo December 1990
Detailed Cornments by R.B. A. Licciardo on T. Murley Closure of OP0 lssues Regarding the Mc Guire Technical Specifications, j
dated September 10, 1990.
]
Question 7c and 7h Table 3.3 5.
Item 2d: Table 3.3-5, Item 3d.
RESPONSE
There is no response to this question.
This is unacceptable.
Action:
Licensee shujld review RCPB valves isolated by the Safe.ty Injec-tion signal to ensure shortest possible closure timer, consistent with any 1
specific analyses using particular valves which shoL d already have been incorporated in relevant T5's.
Such closure titres should be incorporated into the T5's.
2-18 Robert B.A. Licciardo December 1990
]
Detailed Comments by R.B A. Licciardo on T. Murley Closure of OP0 Issues Regarding the Mc Guire Technical Specifications, dated September 10, 1990.
Question 7e, Table 3.3-5, Item 2f:
RESPO!4SE The licensee has provided no response to the licensing bases information i
provided by the writer under Ref A.14 justifying his proposition on this particular issue.
Therefore the writer's position must be considered uncontested and thereby correct.
Therefore:
Table 3.3-5 Items 2f, 3f, and 4f, shall include response times of equal to or < 60 secs, against the item of Auxiliary Feedwater Pumps.
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2-19 Robert B.A. Licciardo December 1990
Detailed Comments by d.B.A. Licciar:9 on T. Murley Closure of DP0 Issues Regarding the Mc Guire lechnical Specifications, dated September 10, 1990.
Question 7j, Table 3.3-5, Item Of.
RESPONSE
The licensee has provided no response to the licensing bases information provided by the writer under Ref. A.14 June 19 justifying his proposition on this particular issue.
Therefore the writer's position must be consi-dereo uncontested and thereby correct.
Therefore: Table 3.3-5 Items 2f, 3f, and 4f, shall include response times of equal to or < 60 secs. against the item of Auxiliary feedwater Pumps.
2-20 Robert B.A. Licciardo December 1990
Detailed Comments by R.B.A. Licciardo on T. Murley Closure of OP0 Issues Regarding the Mc Guire Technical Specifications,
]
dated September 10, 1990.
Question 7o, Table 3.3-5, Item 12 i
RE$r0NSE Acceptable:
The licensee should confirm that with the actual closing time 1
for the sump and RWST valves being 60 secs. shorter than provided for in the sequence described, that sufficient water is ultimately delivered to the containment vessel to establish the NPSH evaluated to be available for all the ECCS pumps.
2-21 Robert B.A. Licciardo December 1990
Detailed Comments by R.B. A. Licciardo on T. Murley Closure of DP0 Issues Regarding the Mc Guire Technical Specifications, i
dated September 10, 1990.
Question 9, Page 3/4 4-2, TS 3.4.1.2
RESPONSE
In his response the licensee has not addressed the need to determine safe-ty limits and thereby TS for restart of a reactor coolant loop in this mode, and thereby is unacceptable.
Restarting an RCP without an adequate reccgnition and Analysis of the pre-vailing conditions and consequences can cause a i,ignificant increase in reactivity, reactor power, and reactor pressure.
The licensing bases for Mc Guire provided for substantially increased Boration concentrations to approx. 2000 ppm in Modes 3-5, to mitigate these potential circumstances; but the existing TS are in def ault in not providing for such Boration lev-els.
Therefore the plant is exposed to potentially undesirable conse-quences if the action proposed is undertaken at this time.
This concern had been recognized as a Generic item under Section 3/4.4.1, G2.6.1 and Listed under Table 4, Question 8a of this Memorandum.
Action: The licensee should be required to re-evaluate for his current TS, or borate to the level required by his existing safety evaluation under Ref. 16 page Q 212-47e before initiating cooldown in Mode 3.
Unless prior boration is agreed to the existing TS petmitting restart of the reactor coolant pumps represent an unanalyzed safety condition and should be im-mediately withdrawn.
The licensee's new propcsal to utilize Abnormal Procedures for Natural Circulation are acceptable but not in the event that the the first priori-ty is to use the RCP'S without prior boration.
2-22 Robert B. A. Licciardo December 1990
Detailed Comments by R.B. A. Licciardo on T. Murley Closure of DP0 Issues Regarding the Mc Guire Technical Specifications, dated September 10, 1990.
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Question 11b, TS Item 3.5 RESMNSE Reviewers have accepted the writer's proposition that these concerns are to be evaluated and it has been decided to do this on a Generic Basis and for incorporation into the New Standard TS.
This issue however remains a Licensing Basis requirement for the Licensee, even though it is to be treated Generically.
2-23 Robert B.A. Licciardo December 1990
Detailed Comments by R.B. A. Licciardo en T. Murley Closure of DP0 Issues Regarding the Mc Guire Technical Specifications, a
dated September 10, 1990.
l Question 11c, TS 3.5
RESPONSE
Reviewers have accepted the writer's proposition that these concerns are to be evaluated and it has been decided to do this on a Generic Basis and i
for incorporation into the New Standard TS.
This issue however remains a Licensing Basis requirement for the Licensee, even though it is to be treated Generically.
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2-24 Robert B.A. Licciardo December 1990
Detailed Comments by R.B.A. Licciardo on T. Murley Cicsure of OP0 Issues Regarding the Mc Guire Technical Specifications, dated September 10, 1990.
Question 12a, T5 3.5.1.1.d
RESPONSE
Reviewers have accepted the writer's proposition and the FSAR is to be updated to reflect this.
Related Set Point methodology will need to be amended to reflect the changes in drif t and channel error, and also the TS insof ar as these amendments result in a change to the related Set Points and Allowable values.
2-25 Robert B.A. Licciardo December 1990
3
]
4 Detailed Comment $ by R.B.A. Licciardo on T. Murley Closure j
of OP0 Issues Regarding the Mc Guire Technical Specifications, dated September 10, 1990.
i Question 12b, TS 4.F.1.1.1.d.1 i
RESPONSE
i Reviewers have accepted the writer's proposition and the IST program, and necessarily the FSAR, is to be updated to reflect this.
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'2-26 Robert B.A. Licciardo December 1990
Detailed Connents by R.B. A. Licciardo on T. Murley Closure of OP0 Issues Regarding the Mc Guire Technical Specifications, dated September 10, 1990.
Question 13, TS 3.5.1.2.d
RESPONSE
The f undamental safety issues of this item were identical to those of Question 126, TS 3.5.1.1.d an6 Question 12b, TS 4.5.1.1.1.d.1 above, and were therefore accepted by the reviewers; however since the equipment has since been removed no related changes to the IST, FSAR and SPM are now necessary.
2-27 Robert B.A. Licciardo December 1990
O o
l Detailed Comments, by R.B.A. Licciardo on T. Hurley Closure of OP0 Issues Regarding the Mc Guire Technical Specifications, dated September 10, 1990.
Question 14, TS 4.5.2.h
RESPONSE
The necessary distribution of ECCS flows to effectively protect against a LOCA is a set of Safaty Analysis Limits which must therefore be recorded inside the FSAR.
l 2-28 Robert B.A. Licciardo December 1990
Detailed Comments by R.B. A. Licciardo on T. Hurley Closure of DP0 Issues Regarding the Mc Guire Technical Specifications, dated September 10, 1990.
Question 17, TS 3/4.7.5
RESPONSE
Reviewers have accepted the writer's proposition and are therefore rea quired to include surveillance requirements in the TS to ensure that the operating train (of the Nuclear Service System is manually aligned to the Nuclear Service Water Pond under icing conditions.
The importance of this is critically increased by the f act that this change-over is manual, not au-tomatic, so that under accident conditions automatic response for the sup-ply of Nuclear Service Water service water is not available within the necessary response times of 65 to 76 secs of Table 3.3-5.
The licensee shall Amend the TS's to include this requirement under either TS Section 3/4.7.4 NSWS or TS Section 3/4.7.5.
SNSWP.
No change to the FSAR is necessary as the commitment remains in the document.
Note:
The 76 secs is outside the Licensing Basis for the plant as described under CINS 180, 181, 182, 192, 195, 204 and 212 and must be evaluated.
Otherwise the plant will be in an Unanalyzed Safety Condition.
2-29 Robert B.A. Licciardo December 1990
Detailed Comments by R.B.A. Licciardo on T. Murley Closure of DP0 !ssues Regarding the Mc Guire Technical Specifications, dated September 10, 1990.
Question 18, TS 3/4.9.1
RESPONSE
The reviewers have not provided a detailed response to the writer's safety evaluation of the need for TS changes on this issue to restore licensing basis protection requirements of safety related engineering and surveil-lance procedures to protect the plant against the boron dilution events.
The writer must therefore conclude that there is no defensible position by the reviewers and the writer's evaludtion remains valid.
Action:
The licensee shall modify the language of his TS to require lock-ing of the valve #1NV 250, and to' verify closure of the valves IRV-171A and INV-175A.
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l 2-30 Robert B.A. Licciardo December 1990 i
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i COMMENTS ON THE REVIEW OF GENERIC DP0 ISSUES OF THE R.B.A. LICCIARDO PART 3:
MCGUIRE TS REVIEW OF 1984 (REF. A.1)
PART 3.1
SUMMARY
j 4~
The total number of items identified for generic consideration by a minimal set of various entities is 240.
The total number of necessary additions to the NSTS is 207 and of this count many are also included in the WSTS under the same item numbers (CINS)
The total number of CINS impacted by both changes to the ETS and or the WSTS is 170, and would represent the total impact on the Existing Tech. Specs, alone for i
the MCGUIRE UNITS and which would have protected the plant against the Mid-Loop loss of Residual Heat Removal Cooling at both the Diablo Canyon and 3-Votgle units In-conclusion the level and quality of the NRC review in the area of Generic concerns is Unacceptable as a Peer review for the writer's McGuire TS Review of Ref. A.1 PART 3.2 DISCUSSION OF MINIMAL SET OF GENERIC ACTIVITIES AND ACTIONS ARISING FOR VARIOUS ENTITIES, AND DERIVING FROM THE R.B.A. LICCIARDO MCGUIRE 4
TS REVIEW OF 1984 (REF. A.1). REF. TABLE 3.1 (A SUBSET OF TABLE 1)
This table 3.1-lists the total set of generic actions undertaken by various entities since the preparation of the McGuire TS Review. These include Generic Letter 86-13 by the NRC To All Power Reactor Licensees And Applicants With-Combustion Engineering and Babcock and Wilcox Pressurized Water Reactors, A Westinghouse letter to All Westinghouse Licensees on the subject of the Number Of Reactor Coolant Pumps In Mode 3. A Westinghouse Owners Group Action on LOCA in-Mode 4;. items submitted for. inclusion in the Westinghouse Standard TS by the former Reactor Systems Branch; particular Generic studies by W and the NRC, r
Particular-TS submittals by W through Licensees on overpressure protection in Modes 3 through 5, and other items as identified in the Table Nomenclature, j
In table 3.1, the impact of generic issues on the NSTS is-identified in two c
general ways a)-
The total number of items identified for generic consideration by a-minimal set of various entities is 240 b)-
The total number of necessary additions to.the NSTS is 207. Of this count j
many are also included in the WSTS under the same CINS and which are not separately accounted for c) -The impact on the NSTS arising from the NRC reviews where the results 'are represented by specific Reviewers conclusions is represented by the locator NSTS. 'These total 17, 3-1
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d) The necessary supplement to the Reviewers specific conclusions as concluded by the writer's comments in Part 3.1 of this memo are identified by the locator NSTS+ and total 190 additional items, e) Detailed analyses will show that an additional 4 items are added to the WSTS only and not accounted for in this particular analyses. A cross check with Table I will show this result. This gives a total of 211 CIN'S for the NSTS and the WSTS PART 3.3 DISCUSSION OF MINIMAL SET OF ACTIONS ON THE EXISTING TS AND WESTING-HOUSE STS ARISING FROM VARIOUS ENTITIES, AND DERIVING FROM THE R.B.A, LICCIARDO MCGUIRE TS REVIEW OF 1984 (REF. A.1). REF. TABLE 3.2 (A SUBSET OF TABLE 1).
The total number of items in this list is 170, and this represents the total impact on the Existing Tech. Specs, alone for MCGUIRE, or the ETS through the impact of the WSTS, or the WSTS alone, both from the action of the A. Thadani Letter discussed in Table 2.1 and also the impact arising from other actions including the ongoing generic actions of tha other entities which would be incorporated into WSTS or ETS before adoption of the NSTS. This represents are larger Set than discussed in Part 2 The total no. of actions involving the WSTS directly, and the Westinghouse TS through actions initiating in the Existing TS (ETSW OR ETS.W) is 100. This limited set is occasioned by earlier reviews before the advent of the NSTS in which references were necessarily made to that document iri generic studies:
And also which evolve out of multiple Plant Specific Actions such as the "The Number Of Reactor Coolant Pumps Operation In Mode 3".
The impact on the WSTS arising from specific Reviewers conclusions is repre-sented by the locators ETSW or ETS.W. or WSTS.
These total 25. The necessary supplement to the Reviewers specific conclusions as concluded by the Writer (RBAL), are identified by the locators ETS.W+ or ETSW+ and total 75 additional items.
Reference the detailed Comments elsewhere in this Report.
The table shows that many of these items also become part of the NSTS 3-2
j 4
i PART 3.4 DETAILED COMMENTS ON THE REVIEW OF THE PROPOSED CLOSE OUT OF GENERIC ISSUES BY THE T.H. MURLEY MEMO 0F SEPTEMBER 10 1990 is a copy of the REVIEW OF MCGUIRE TECHNICAL SPECIFICATION to l
i Ref. A.I. This identifies the Congressional item Numbers (CINS) for each of the items of concerns reported to the US Congress Ly the Chairman Nunzio J Palladino of the USNRC by letter of December 20, 1984. is the copy of the T. Murley memorandum to Ref 40 and which is a principal subject of these Comments. These detailed Comments are made against directly against Enclosure 3, sub-enclosure 4 and particularly against each of the items in the sequence in which they are presented in that document. Where no further comment is made the item is not generally addressed:
Reference to this document is essential.
The Reviewer's have not detailed their review of the writer's concerns for CINS 292 through 298 even though these fully evaluate the REACTOR COOLANT SYSTEM-COLD SHUT 00WN, LOOPS NOT FILLED, which is the Diablo Canyon Event of 1987 for which the licensee was completely unprepared because the NRC rejected his concerns outright in early 1983 and when detailed under the current DP0 review in 1984 were again given very low priority even beyond the event itself until the LOCA in Mode 4 event at Braidwood in December 1989 when it then received the first consideration under this accelerated review.
Acceptance of the Writer's concerns in 1984 would have ensured awareness of the event on its Occurrence and complete protection, instead of the severe risk to which the public was exposed. These circumstances also apply to the Vogtle Event under a significantly different set of circumstances, in Mode 6, later in March 1990 which was covered under CINS 399 to 405 which would have protected the plant i
against the event--The staff's comment that none of the Writer's issues applied because none of them concerned Station Blackout reflects the fact that the NRC j
has still not studied these comments 61/2 years af ter their preparation:
The NRC Staff is again invited to read the Writer's CINS references above, and his detailed Comments under Concern 36 A of this review, and prepare a valid safety i
evaluation of his propositions and discover where their problems exist and potentially facilitate an overall improvement in their proposed total level of protection.
It must be said that the level of the Reviewers comments do not represent the level of impartiality that is recessary to ensure that never again will there be serious misjudgment about the importance of writer's early propositions on potenti'111y serious events before they occur as has been manifested for the cases of Diablo Carr 2n, Braidwood and Vogtle. And there are many more events still waiting to happen in the residual unprotected state in which the current NRC Reviewers would propose to leave the NSTS and the WSTS without the provisions introduced by the writer in this Review.
In respect of the 160 items not selected for review in 1984 even after the writer explained their importance in e memo to Denton (Ref. 36), the current staff persists with the 1984 decision even though Reactor Protection in Modes 3-6 is now a major research program and they allegedly have reviewed for the necessary protections and generic items in modes 3 through 6.
Reference to Part 1.2 discussion will reveal that 88 out of actually 178 open items were identified for evaluation as primarily generic concerns leaving a residual of 86: Furthermore, a detailed check of the 86 residual items by the groups into which they remain show the items are valid and should now be considered; the 3-3
O a
writer's safety evaluation of these items in his 1984 TS Review to Ref. A.1 remain unchanged The NRC review to Ref. 40 has suffered from a number of deficiencies, and these are revealed above and further detailed in the Comments which follow. Many of these are common to those outlined under Part 2.3.1 but with particular impor-tance in certain areas for Generic Issues. Summarily, they include the following:
In considering a set of disparate events with a particular common safety characteristic Reviewers have oversimplified the evaluation by focussing on only one event and in a manner from which generalized invalid conclusions are derived for the remaining events In considering single events, the Reviewers have considered a less than minimum partial set of the information required for the safety analyses of the events and have thereby f aulted in their safety evaluation and in specifying a less than minimal set of such information for inclusion in the TS's.
An item of particular importance here is the set of process parameters used as the starting bases for all T&A's in all Modes of operation 1-5, and not only Modes 1&2.
The reviewers continually propose positions outside the licensing bases for the McGuire units and which are thereby invalid; have misread the Writer's evaluations and have also made evaluations based on faulted knowledge of the Protective logic.
For some concerns the licensee has made faulted statements requiring additional cycles of evaluation.
evaluation cycles for correction. Questions airected to resolving specific issues have been ignored.
There is a marked lack of capability in the necessary detailed Nuclear Engineering of Protective systems, to meet Regulatory Requirements There is a marked and thereby very serious lack of capability in the understanding, and thereby establishing of, the detailed safe operation of the reactor under normal operating conditions, to ensure licensing basis safety under T&A conditions.
In reviewing events in Mode 3, 4, and 5, the Reviewers have used invalid information thereby reaching f aulted conclusions:
Further in the process they have been unable to detect basic faults in this invalid information which should have made them aware of its severe limitations In Modes 5 and 6, the Reviewers have shown a singular lack of knowledge of related licensing bases requirements in evaluating and analyzing for the consequences of T&A's.
The reviewers have also shown a marked lack of the necessary detailed know-ledge of the related Regulatory Requirements for all Protective systems including manual operations thereto for the reactor and the facility in general in Modes 3, 4, 5 & 6, end the detailed nuclear Engineering necessary to achieve these. This has lead to speculation on important features of events and the necessary protective responses by operating staff and including the availability of protective equipment and the 3-4
i preparation of procedures including the TS.
These circumstances are completely Unacceptable within the licensing Bases for ensuring Public Health and Safety A very serious fault and particularly endemic in considering T&A's in Modes 3,4,5 and 6, is the perception that because a particular event may result in larger margins to safety than calculated for the bounding event of the licensing bases, that protection is not needed; and furthermore this conclusion is niade without a safety evaluation of what then happens to the unprotected plant for which the event is now not mitigated because it is now unprotected and thereby not terminated.
In conclusion the level and quality of the NRC review in the area of Generic concerns is Unacceptable as a Peer review for the writer's McGuire TS Review of Ref. A.1 3-5
DetaileG Comments by R.B. A. Licciardo on T. Murley Closure of OP0 !ssues Regarding the Mc Guire Technical Specifications, dated September 10, 1990.
CONCERN 9A, QUESTION Be, TS 3/4.2.5:
AVAILABILITY OF RCPs' DEPARTURE FROM NUCLEATE BOILING (DNB) TS General Comment:
These concerns derive from TS Section 3/4.2.5 under CIN'c 67 to 73, and unfortunately, Reviewer's comments are made outside the context of related concerns of the writer in Section 2.1.1 under CIN's 1-7 and TS section 3/4.4-9 CIN'S 306 to 309.
Reference to these other sections will provide the answer to a number of the Reviewers' comments.
Concerning Resolution I First para:
It was the writer who first proposed the proposition from which the first comment by the reviewer is made.
Under a resulting action by RSB it became a generic issue from which ultimately the TS criteria were developed:
Ref, the previous ref's and also Table 2, CINS 6, 7, and 8.
Second para.:
Reviewers comments are incorrect.
The W reactor programs an indicated Tavg against Thermal power level from and at Zero power in Mode 2 to Maximum Licensing Basis power in Mode 1:
Ref CINS 67-69.
Safety analyses uses these programmed values at Zero power and maximum licensing basis power in conjunction with positive and negative errors in the related measuring instru-mentation to provide the upper and lower safety analyses limits in calculating the consequences of Licensing Bases Transient and Accidents Analyses in the approach of the plant responses to the safety limits of multiple criteria:
and not only DNB as related by the Writer in reference A.1, Page 16, Section 3.4.2.5./
Evaluation / Item a).
Further, these evaluations are also undertaken at interme-diate power levels such as P-10 and P-8.
In general, for nuclear systems the margin to the safety limit of a given criterion is not measured directly, but necessarily calculated using a No of variables for parameters which themselves are not the safety criterton.
The critical values of the process parameters themselves therefore do not have a safety limit but have a safety analysis limit, t
and likewise a limiting safety analysis limiting value, and a safety analysis Set point.
And for Plant Protection these are generally identified in the Set Point Methodology for the particular Nuclear Unit.
And therefore these programmed values must be included in the TS under 10CFR50.36(c)(1) with the necessary safety analysis limits, limiting safety analyses system settings, and set points, which have been the CINS of concern identified above in his OPO.
These parameters must thereby ultimately include related values for Tavg, pressurizer pressure and level, and steam generator level, as are discussed in the Writer's DPO.
Allowable values for expected eM ft should be established within the Limits of
(
Total Error (Ref. CIN 1), and indeed the latest MC GUIRE TS incorporate such i
limits for (maximum) Tavg.
ano possurizer pressure,and these become the related limiting safety analysis system uttings.
The sub-item a) of this para. is incorrect.
The reviewer should detail read j
the licensees CINS referenced a W e for the related references.
Process Set Points do not need to generate a trip; they represent necessary setpoints for operation of the plant under st @le normal operating conditions and are expected to vary slowly only witH n the ability of the related control systems l
l 36 Robert B.A. Licciardo December 1990
-a
. - - -=-_--
Detailed Comments by R.B. A. Licciardo on T. Murley Closure of OP0 Issues Regarding the Mc Guire Technical Specifications, 1
dated September 10, 1990, to sustain them:
If they cannot be sustained within the allowable values (the limiting safety analysis system settings), then the plant must be shut down as otherwise the calculated safety upon which safety of the plant has been estab*
lished and licensed can no longer be validated.
Alternately, if there is a transient or accident causing rapid change to unsafe values, the reactor pro-tection system will protect the reactor through the use of the overpower and I
overtemperature Delta T trips and or other trips of Table 2.2.1 and in accor-dance with the new set points and limiting safety analysis system settings cal-culated to be required for such a Transient or Accident from the related bound-ing event.
References to these and related graphical representations are discussed under CIN 2.
The sub-items b) and c) are incorrect statements in the light of the previous paragraph.
==
Conclusion:==
The writer's concerns under the complete set of CINS reference above remain valid, and the related clarifications and necessary amendments to the TS should be incorporated in both the New Standard TS (NSTS) and the Plant Specific TS (PSTS).
Concerning Resolution II The Writer's comments for Tavg under Resolution I above also apply to the Rt viewers' comments here for pressurizer pressure, since all the determinant cb cumstances are the same.
The reviewers acknowledge that pressurizer pressure is also a an important process variable necessary to protect the integrity of the physical barriers that guard against the unconditional release of radioac-tivity and thereby it must be included in the TS under 10CFR50.36(c)(1) with the necessary safety analysis limits, limiting safety analysis system settings, and safety analysis set points which have been the CIN5 of concern identified above in his OPO.
In this case, if there is a transient or accident causing rapid change to un-safe values, the reactor protection system will again protect the reactor through the use of the overpower and overtemperature Delta T trips, the pres-surizer pressure trip, and or other trips of Table 2.2.1, and again in accor-dance with the new set of set point and limiting safety system settings calcu-lated to be required for such a Transient or Accident from the related bounding event.
References to these and related graphical representations are discussed under CIN 2.
Concerning related para. 3:
Reviewers should reference CIN 11 which identifies and discusses the design pressure for the mechanical design of the reactor coolant system and its internals at 2485 psig.
2250 psia from table 4 in which section the thermal hydraulic evaluation of the reactor core is evaluated must be taken as the instrumentation set point for control of the pressurizer pressure under normal stable operating conditions and which is used to calcu-late upper and lower safety analyses limits (by the application of specified instrument error corrections) from which plant safety is calculated in the same manner as for Tavg above:
Except that pressurizer pressure remains constant from Zero power to traximum licensing basis power, unlike Tavg.
Under these conditions the TS Set point should at 2235 psig instead of the value of 2215 psig used in the TS at that time.
3-7 Robert B.A. Licciardo December 1990
Detailed Comments by R.B.A. Licciardo on T. Hurley Closure of OP0 Issues Regarding the Mc Guire Technical Specifications, dated September 10, 1990.
Concerning the last para. and the calcul:,sion of the safety analysis limit, when conservative methodology is used for T&A's, safety analysis limits, for both upper and lower values, are calculated from the steady state instrument reading setpoint as described above.
The comments by the reviewer are thereby incorrect.
Concerning Resolution-III The meaning of the values on Fig. 2.2.1 have been concerns by the writer under CINS 1, 2, & 3 and which have never been satisfactorily answered by multiple reviewers:
Its oriO n with respect to the licensing bases for verification for inclusion in the TS is not evident as it is no presented it that form in the FSAR.
And likewise current Reviewers have also chose not to respond.
By logic if Fig. 2.1.1 represented allowable normal operating conditions for satis-factory protection against all licensing basis T&A's then programmed values of pressurizer pressure should be included.
If the set-points for pressurizer pressure as discussed herein are to be repre-sented on this Figure and which was originally proposed for consideration by the writer under See. tion 3/4.2.1, Evaluation a) and CIN 72, the only licensing basis that exists is ae a constant pressurizer pressure of 2235 psig over the range of thermal power from zero to the maximum licensing basis rating and which should be labelled as Acceptable Operation under steady state operating conditions.
The current TS which do not provide for this would be outside the existing licensing basis.
There is no licensing basis for the implied proposition that steady state opera-tion anywhere inside the regime of " acceptable operation" will give Acceptable Responses under licensing basis T&A's.
This proposition is false. Without the further clarification requested by the writer in his OP0 under CINS 1&2, it appears that this figure represents potential safety limits for reactor opera-tion under steady state conditions only, and there is no licensing basis evalu-ation to show that transients and accidents occurring whilst operating under this broad range of conditions will give Acceptable responses.
The only steady state conditions from which acceptable responses have been calculated to be safe are the programmed values of Tavg and Pressurizer Pressure (and level, and steam generator levci), versus Power Level as has already been described and it is these values which need to be included in this figure:
Any other plant status would place the plant in an Unanalyzed Safety condition.
Concerning the Operability of the Pressurizer:
CIN 307 shows that pressurizer (water) level programmed with power is also a parameter requiring inclusion in the TS for the same reasons as has been dis-cussed above for Tavg and pressurizer pressure.
For licensing basis T&A's, Pressurizer operability is established only when it is capable of maintaining a pressurizer pressure at its set point value (of 2235 psig) and the programmed set point values of pressurizer water level, dur-ing steady state operation, and this depends not only on the water volume and heaters but the the complete set of subsystems contributing to the maintenance of these values.
Thereby the performance requirements remain the only valid 3-8 Robert B.A. Licciardo December 1990
Detailed Comments by R.B. A. Licciardo on T. Murley Closure of OP0 1ssues Regarding the Mc Guire Technicci Specifications, dated September 10, 1990, bas for the related LCO's.
By logic if Fig. 2.1.1 represented allowable normal i
operatinn en.'11tions for satisfactory protection against all licensing basis T&A's then programmed values of pressurizer level should be included and labelled as Acceptable Operation under steady state operating conditions.
Note that the operability of the PORV's on the pressurizer are also a require-ment for protection against Steam Generator Tube Rupture and thereby may need to be considered as part of the operability of this item (pressurizer).
This should be an additional item for consideration under the related CIN 309.
3-9 Robert B.A. Licciardo December 1990
Detailed Comments by F..B.A. Lieciardo on T. Murley closure of DP0 Issues Regarding the Mc Guire Technical Specifications, j
dated September 10, 1990.
CONCERN ISB, QUESTIONS Ba, 8b, 8c, 8d, and 8e.
TS 3/4.4.1*
REACTOR COOLANT LOOPS AND COOLANT CIRCULATION.
q RBAL COMMENTS:
These are made for each of the main headings of this group:
4 Issue:
i The aim of
" review in simplifying the writer's concerns has grossly mis-represented sensing bases for Mc Guire in this Mode 3 and following Modes 3,4,5 and 6.
...,equently the reviewers' comments are invalid in their repre-4 J.
sentation of the writer's concerns.
As an example, the Mc Guire licencing basis requires Special Boration Procedures and related protective measures on entering Mode 3 through to cold shutdown 4
(Mode 5) to substantively minimize the necessary protective requirements for Acceptable levels of protection against all appropriate T&A's:
Under these conditions the boron concentration on entering Mode 3 is increased to the value required in Mode 5.
Under these circumstances shut down margin in Mode 3 is substantially increased (over that required by the STS) such that if the reac-l tor is already tripped at the beginning of the event the return to nuclear pow-er event is substantively ameliorated.
Huwever the reactor must be already tr_ipped, and the reduction in absolute pressures in these modes, albeit with decreased reactor coolant system temperatures, serves to substantively reduce DhBR margins or increase fuel damage.
Furthermore, for some occurrences the event itself must be terminated, even though the reactor is tripped, otherwise it would proceed beyond the limits of normal protection in Modes 1 and 2, and 4
thereby lead to a severe accident.
Note that if the reactor is not initially re quired to be tripped by TS, and necessary safety related reactor and Engi-nenred Safety Features trips are not incorporated into the TS in these Modes, the n any T&A still has the probability of generating a severe accident.
By comparison, the STS upon which the licensee's TS was based, was developed primarily to assure adequate decay heat removal capability in these Modes 3,4,5 and 6 without consideration of the need to protect against any Transients or Accidents.
Consequently it is absent any Special Boration Control Procedural requirement and is virtually absent any safety related protective trips and thereby the plant remains completely unprotected with a high probability of a severe accident arising from the Occurrence of any Transient or Accident.
Question Ba:
OCCURRENCES WITH RAPID REACTIVITY INCREASE The first comment limits these Occurrences only to the Uncontrolled Rod Cluster Control Assembly Bank Withdrawal from Sub-Critical Condition, whereas there is a set of at-least seven Occurrences, each with different characteristics.
-This comment in the second para. attributed to the Writer is made inside the context of the provisions of the existing TS for Mc Guire which does not con-form to the Mc Guire FSAR licensing basis, but to that of the STS with,all of its related deficiencies.
3-10 Robert B.A. Licciardo December 1990
~
O e
Detailed Comments by R.8. A. Licciardo on T. Hurley rs ure of OP0 Issues Regarding the Mc Guire Technical Sped 4:nions, dated September 10, 1990.
Question 8b:
STEAM LINE BREAKS:
OCCURRENCES This statement on the resulting impact on safety margin it N h inside the con-text of the existing FSAR with Special Boration Control t w m eific require-ments for automatic and manual protective actions to prote:" ete reactor against the occurrence.
And since as previously describta 1'tse FSAR require-ments are not included inside the STS upon which the Mc G m q 'S was based they cannot be used as a valid basis for the Reviewers' conti n M t. which as clearly stated in his Report under CIN 247 are based on tat F rcumstance of the proposed (current) TS and its deficiencies in protective Nt' ns.
These are all fully discussed under CINS 244, 245, 246, and 247, to W ch no response has been made by the reviewer's.
The Reviewers, conclusiont u t v erefore unac-ceptable and the Writers' concerns remain valid.
If the Reviewers wish to take advantage of the Special Boration Control provi-sions of the Mc Guire FSAR, then CINS 41-66 and 355-362 rust ce addr(ssed to-gether with the additional provisions for initiating manut' and nutomatic action as described in the licensing basis for Mc Guire and e m ided for elswhere in the DPO.
Question 8c:
LOSS OF PRIMARY COOLANT:
OCCURRENCES It should be noted that the positions taken by the writer are not Assertions but propositions deriving f rom available information whitt seeking further evaluation and proposals from the licensee to finalize a safe position:
They would become the basis for further licensing action in the event further At-ceptable safety evaluations were not obtained.
i Question 8d:
OCCURRENCES CAUSING AN INCREASE OF RCS TEMPERATURE l
l The Reviewers have neglected to mention the most important consideration in including events which are normally licensing basis events from Rated Power, as potentially significant events f rom Zero Power.
As the writer has explained from his review to Ref. A2., CINS 257 through 261:
"Those events causing an increase in RCS temperature are of con-cern because of the potential influence of the positive moderator temperature coefficient resulting from the increa6ed boron concentration" l
And further:
Except for item b; all these events are licensing bases events from Rated power, and not zero power, so that their importance l
l would normally be minimal except for the positive moderator Temperature Coefficient and the complete lack of Safety Related l
Trip protection proposed with the Reactor Trip System Instrumentation TS."
l 3-11 Robert B.A. Licciardo December 1990
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Detailed Comments by R.B. A. Licciardo on T. Murley closure j
of DP0 Issues Regarding the Mc Guire Technical Specifications.
[
dated September 10, 1990.
Question Be:
AVAILABILITY OF REACTOR COOLANT PUMPS I
Again the reviewers have represented the writer'r position in a very simplistic and inadequate manner in respect of all the important related significant con-j siderations.
The writer summarized the position as:
E
" Occurrence II, III, and IV Events in MODES 3,4 and 5, can result in retur..s to power with high peaking coefficients requiring effective reactivity control i
and/or reactor core flow for RCS protection, including DNBR, at the very sub-i stantially reduced pressure levels in the loop (2250 psig to 425 psig and less).
Concomitant decreases in RCS temperatures are beneficial, but the im-portance of RCS pressure may be dominant.
Acceptable RCS protection therefore requires RCS flows which are substantial, and/or effective reactivity control j
including combined action to limit potential reactivity excursion.
At this time, with the proposed TS, 4 RCS loops (with increased Reactor Trip Protection) would be required at entry into and during Mode 3 to meet the re-quirements of just the Licensing Basis Events From Zero Power.
In Mode 4, op-eration of 4 RCS Loops, whilst in RHR, may be undesirable because of the substantial additional burden on the RHR system; so, nonoperability of all RCPs q
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 Aux-iliary Feedwater systems, closure of Main Steam Isolation Valves, and Boration Control measures additional to those included in the propose TS.
An additional available alternate action is to use, within Mode 4, a minimum set of RCS pumps (and loops) as established by Safety Analysos, to cool the plant down to effec-tively 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 writer must conclude that in excluding this summary representation of the i
writer's concerns that-the Reviewers, are not capable of the quality of the review that was called for in the agreement to subject the writer's comments to i
what was to be effectively a peer review.
Comments on the " Resolution" The new STS now recognizes in a very limited manner the writer's concerns but i
because of the inability of the reviewers to recognize, evaluate and consider for incorporation into the TS's all the elements fully described in the writ-er's review that are important to Acceptable Plant-Protection in accordance with existing licensing basis requirements, the plants adopting such TSs wil'1 3
remain unprotected against potentially severe consequence in Modes 3, 4 and 5, by being exposed to a complete set of Unanalyzed Safety Conditions.
In representing the un-referenced Westinghouse positions to writer's reference A.13, the Reviewers have displayed a remarkable degree of non-conformance to i
Regulatory Requirements in representing a letter has having Acceptable Regula-
-tory Positions when to the best of the writer's knowledge it has never been reviewed and formally accepted by the NRC as a Topical Report.
3-12 Robert B.A Licciardo December 1990 l
l 1
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o Detailed Comments by R.B. A. Licciardo on T. Murley Closure of OP0 Issues Regarding the Mc Guire Technical Specifications, i
dated September 10, 1990.
Furthermore in reviewing and applying the Westinghouse letter to Licensing Bases protection in these mod >. 1-G the reviewers have not shown the capabili y of an Acceptable appraisal of its ability to protect the core as required in the licensing basis for the facility.
And in spite of the fact that all these circumstances and conditions and related necessary considerations are presented in the Safety Evaluation Report known as the REVIEW OF MC GUIRE TECHNICAL SPECI-FICATIONS to Reference A.1.
The writer would be pleased to provide a licensing hasis Safety Evaluation Re-port on the Westinghouse Letter and offered to do so after receiving a copy of the document in July of 1984 in response to the Writer's earlier Safety Evalue-tion Report on the Mc Guire TS which was given in an Unauthorized Non-Regulatory manner to Westinghouse.
However the writer's request was refused when he ex-pressed concern about serious deficiencies in the effective representation of the report leading to potentially unsafe conditions for the reactor.
Unfor-tunately, since then it han apparently been represented as a satisfactory basis for licensees to make appropriate representations in respect of Amendments to their TS and now the New Standard Technical Specifications (NSTS) for which it is seriously deficient.
Examples will suffice:
W proposes that consideration of the Uncontrolled Rod Withdrawal Event in Modes 4&5 is unrealistic.
This is not the licensing basis for Mc Guire where no such restriction was acceptable:
It was the same reason given by the NRC Staff in rejecting the mid-loop operating event first identified by the writer inn 1983 from this Mc Guire TS Review and together with oth6: concerns eventually re-suited in this DP0 Review.
This is Unacceptable.
The only reason for propos-ing this as unrealistic is that it is extremely difficult to otherwise protect against except in a very simple manner by unlatching the control rods in these Modes.
It should be realized that the consequences of such an unprotected event in these Modes 4&S would be a severe accident.
W proposes that for the Reactivity Insertion Rate for the Uncontrolled Rod Withdrawal event be assumed for 30 pcm/s compared with the licensing basis re-quirement for this event of 75 pcm/s.
Thereby the W proposition is invalid and Unacceptable.
W proposes a bounding condition for the R.C.S. in Mode 3, at 400 deg. F and 2000 psia.
Surely Reviewers, experienced in these reviews, would recognize a more appropriate related bounding condition of 425 psig at RCS Temperatures >
350 deg. F being the RCS condition in Mode 3 prior to entry into the RHR Mode and under which the substantially reduced pressure would lead to a severe acci-dent.
Thereby the W position is invalid and Unacceptable.
What safety related protective system has W described to ensure that the reaca tivity turns around as represented in its submittal.
The only safety related l
reactor trip available for this purpose on Mc Guire Units and preceding plants is the Power Range Neutron Flux Trip-Low Power Set Point, and the Mc Guire TS l
and the STS at the time of this DP0 review did not require tnis trip to be op-erable in Mode 3 (as well as Modes 4, and 5):
And this has not been t caution-l ary important advisory in the W presentation.
So the reactor could be subject to a severe accident from this unprotected event and this is Unacceptable.
3-13 Robert B. A. Licciardo Decenber 1990 1
l, Detailed Comments by R.B.A. Licciardo on T. Murley Closure of DP0 1ssues Regarding the Mc Guire Technical Specifications, dated September 10, 1990.
Referring to the second last para. of the " Resolution", the writer finds no difference with the Reviewers, on his representations on the question that Re-actor Coolant Pumps operating throughout the critical phase of a LOCA will re-duce the calculated related peak clad temperature and that thereby the coast-down of tripped pumps or pumps which have Lost Offsite Power is similarly beneficial but with lesser effect:
Refce. A.2 page 64.
The point at issue of the Reviewers, seems to be the writer's statement that pending further analyses, these considerations warrant the required operation of 4 p aps to ersure ade-quate protection against LOCA'S down to 425 psig/350 deg. F.
This i: based on the fact that the negative core flow rate occurring on the loss of the pumps during a LOCA would be consistent with that of the 4 pumps used in W ECCS ana-lyses.
If only 2 pumps were operating in this Mode instead of 4, the negative flow rate and its beneficial effect would be reduced and thereby be outside the licensing basis, and result in a higher calculated Peak Clad Temperature which would thereby be Unacceptable.
The above examples and many other features of the original W presentation to Ref. A.13 make it inappropriate to spend further time on the Comments of the Reviewers who have not been able to evaluate for any of the significant defi-ciencies in its use as a proposed Topical Report let alone an Unauthorized Guide.
And on these considerations the writer finds the conclusions by the Reviewers in their last para. of the Resolution to Question ISB to be unacceptable, in conclusion on this particular issue, all the writer's CINS associated with this particular CONCERN ISB whicn are widespread through-out his OP0 have been validated and it has initiated a series of events with far reaching and wice-spread implications.
In this respect the writer references the reader to TABLE 3 in which a minimum set of its w'despread effect on TS CINS is identi-fied alongside the code WL for Westinghouse Letter to utilities and running l
parallel with GL, the Generic Letter from the NRC.
And added to that now must be the major activity initiated by the Office of f.uclear Reactor Regulation early 1990 and the purpose of which is to now formally study Reactor Protection in these Modes 3 through 5 and 6.
The Writer trusts that the results from these developments will now recognize the importance of the Mc Guire FSAR in establishing a Licensing Basis for pro-tection against all appropriate Transients & Accidents in Modes 3 through 5 and 6 and together with the results from the writer's review establish and confirm an NRC Policy in this matter.
l l
l l
l 3-14 Robert B.A. Licciardo December 1990
Detailed-Comments by R.B. A. Licciardo on T. Murley Closure of DPD 1ssees Regarding the Mc Guire Technical Specifications, dated September 10, 1990.
CONCERN 18A:
QUESTION 10, TA Page 3/4 4-3.
REACTOR COOLANT SYSTEM--HOT SHUT DOWN, Comments on the Sunmarized Issue:
The Reviewers have eliminated two of the most important sets of requirements from their Summary which are 1) the totality of protective elements needed to protect against T&A's in this plant status under the related RCS loop operabil-ity circumstances and 2) The Reactor Coolant Systi.m (RCS) conditions for which these events were evaluated to ensure their inclusion in the TS's as LCO'S with related Set Points and allowable values.
In so far as these were tiot-addressed in a fully Acceptable manner under the previous Concern 15.B it remains unacceptable for these circumstances and these are detailed in his ref A.I. under CINS 275 through 285.
Since the Reviewers have no comment on these concerns the writer records these as having been reviewed as Acceptable.
The Writer did not address the single failure of a motorized valve arising from the a loss of offsite power.
The passive failure of the valve, independent from that of loss of power Jupplies is a specific licensing basis for this facility.
In this and many other respe:.ts the Reviewers have not responded to the de-tailed requirements for the failure circumstances which were taken from the existing licensing bases for the Mc Guire Units.
The Reviewers thereby wish to create a new licensing bases, and, that is not the purpose of this review.
Comments on " Resolution" For the purposes of-Decay Heat Pemoval Only:
Whereas the MC Guire Units have a single RHR suction line containing two Reactor Coolant Pressure Boundary Isolation Valves, this occurs-in a small number of W units and is'in contravention of Regulatory Requirements and without the neces-nry Formal Exemption which is required under these circumstances.
The passive failure of the valve, independent of power supplies is a specific licensing basis for this facility.
The alternate argument now being used for the NSTS was proposed during licensing of the facility and was Unacceptable.
Regulations require a normal safe shut-down to cold shut down conditions during a Category'1 Seismic event together with a complete loss of.offsite power and the worst single failure.
In the event of-failure of the single valve as discussed, the licensing basis requires return to the use of.the steam generators and under loss of offsite power (LOOP) conditions these would then be required to operate under natural circulation conditions.
Providing onsite control and instrumentation power was available to at least one of these steam generators and natural circulation capability provided adequate heat removal capacity to prevent severe damage to the core, then that would be acceptable:
However if the blockage of the one 3-15 Robert B.A. Licciardo December 1990
l Detailed -Comments by R.B. A. Licciardo on T. Murley Closure of DP0_ Issues Regarding the Mc Guire Technical Specifications, dated September 10, 1990.
-in-line valve was caused by a inadvertent signal caused by a fault in one of safety related power supplies to these Reactor Coolant Pressure Boundary Valves then it must be presumed that the same supply could fault the instrumentation and control systems on the same electrical division supplying Steam Generators (SGS) and so render those-related units inoperable and thereby unavailable.
Since these two valves control two trains they must capable of actuation by either division and therefore a failure in either division could cause its loss:
Unfortunately, this thereby means that steam generators on either divi-sion may be lost (the value of independence has been lost) so that at least two steam generators, one each from separate electrical divisions, must be available under these circumstances and only for the case of those facilities with the common suction line.
The availability of offsite power makes no dif-ference to this conclusion; except that for the LOOP case the licensee should ensure by analyses that sufficient decay heat cooling capacity with one steam generator otherwise two SGS on each division would be required.
These circumstances show that for the nuclear facility with the common suction line to the RHR system, in the event the RHR system is lost, then the alternate use of the RCS loops requires that at least two stean generators, one each from different safety related power divisions must be operable, and of course 2 RCP'S when offsite power is available.
It remains difficult to perceive how one " inoperable system" of two parallel RHR systems sharing common RCPB valves will not potentially impact the remain-ing system, in all potential single failure situations.
For example air induc-tion into one system could also affect the operability of the second system ; a situation in which one RHR pump may be removed would require additional RCPB valves to isolate that system from the operable system and these are not pro-vided. What is the prescribed status of the power supplies and related logic, both AC and DC, which ensures that the RCPB can be isolated automatically in the event this is required and how is this impacted by the fact that these valves also have a logic protecting the RHR from inadvertent overpressurization from the RCS.
And the requirement of the Standard Review Plan that requires that failure of a valve shall not cause any valve to change its position.
And the fact that in its RCPB isolation function, this valve combination should-automatically go to the protected position of being closed in the event of fail-ure of power supply.
The Writer concludes specific inoperabilities would have to be defined to validate the oroposed TS in this matter.
The writer notes that the reviewers have not spoken to the need to ensure that each of the cooling systems required by the TS are required to be powered from separate onsite safety related power divisions (including related DC an AC safety related power supplies for instrumentation and control) to conform to Regulatory Requirements.
Therefore the NSTS proposal that any combination of RCS and RHR loops ce Operable, irrespective of power supplies, is invalid.
Ref.
CIN 287 of Ref. A.1.
CIN 286 of the writer's review (Ref. A.1.) shows that if water solid operation in Mode 4 is to commence at <= 300 deg. F then two_ independent cooling loopt are required to be in operation to prevent an overpressurization event on failure l
of one only operating system.
Further, at this time the writer is unaware of l-any safety analyses evaluation of the adequacy of the existing Low Temperature j
Prot?ction system to mitigate the consequences of such an event; therefore any-3-16 Robert B.A. Licciardo December 1990-I
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.m,
O e
Detailed Comments by R.B.A. Licciardo on T. Murley Closure of DP0 1ssues Regarding the Mc Guire Technical Specifications, j
dated September 10, 1990.
TS permitting such a circumstance would result in an Unanalyzed Safety Condi-tion, and therefore would be invalid.
For The Purposes Of Protection Against Transients and Accidents In Mc A larger number of pumps than one (1) is required unless special mee taken to otherwise mitigate the consequences of these occurrences er requirements will be the determining parameters for the minimum numt.
required to be operating in Mode 4.
Reference our earlier " Comments Summarized Issue", and the original Ref.A.1, CINS 221 through 262, t Comments on second para.:
It is invalid and a thereby a violation of Regulatory requirements to exclude from the NSTS, the safety related Atmospheric Dump Valves (ADV'S) from the TS in Mode 4 since the Regulatory Requirement is to be capable of normal safe shutdown to cold shutdown conditions under a loss of offsite power together with an associated Category 1 Safe Shut Down Earthquake and the worst case sin-gle failure which includes the loss of one electrical division.
The alternate methods of final heat rejection discussed by the reviewers are purely specula-tive in their availability and performance and reflect an approach which is contrary to the regulatory requirements of required safety analyses and 10CfR36, and because of this writer does not provide a detailed response.
The proposed position is Unacceptable.
This subject is also fully discussed under Concern 30A, under the alternate description of " STEAM GENERATOR POWER OPERATED RELIEF VALVES".
Further, there is also a non-safety related function for the ADV's to perform in protecting the plant from a return to power transients in this Mode, in con-trolling energy release prior to the potential lif t of the first SG safety relief.
Arising from the above, the need for operability of the SG safety valves in this Mode has also been raised in the DP0 under CINS 299 to 305 and the review-ers again have not reflected this in the necessary TS.
The Reviewers positions on these necessary protections against T&As in Mode 4 are a reflection of their inadequate evaluations under CONCERN 15B above.
It should be realized by Reviewers that these Occurrences can occur and unless they have been analyzed not to show unacceptable radiological releases under the unprotected plant states they propose, then they remain an UnaMlyzed Safe-ty Condition.
Ultimately they must be protected to that level required by li-censing bases allowable offsite doses otherwise severe consequences for the plant can result.
Comments on para. 3:
Following on the protection requirements just discussed under para. 2, the TS must also ensure the necessary surveillance requirements the purpost of which is specified in 10CFR36 (a)(3) as requirements relating to test, calibration, or inspection to assure that the necessary quality of systems and components is Nintained, that facility operation will be within the safety limits, and that the limiting conditions of operation will be met:
The writer does not see the 3-17 Robert B.A. Licciardo December 1990
Detailed Comments by R.B. A. Licciardo on T. Murley Closure of DP0 Issues Regarding the Mc Guire Technical Specifications, dated September 10, 1990.
necessary responsible evaluation of his proposals for revised surveillance as being undertaken in a manner reflecting the substance and importance of these requirements, as there has been no response to the specific deficiencies in the TS identified by his concerns.
The position proposed by the Reviewers is Unac-ceptable, and thereby the Writer's concerns are valid.
l I
3-18 Robert B.A. Licciardo December 1990
o Detailed Comments by R.B.A. Licciardo on T. Murley Closure of OP0 Issues Regarding the Mc Guire Technical Specifications, dated September 10, 1990.
CONCERN 19A., QUESTION 8.
COLD SHUTDOWN (MODE 5) WITH LOOPS FILLED Concerning the para. " Issues":
This paragraph has not summarized the principal parameters determining the TS requirements in this Mode including the licensing basis requirements for the Mc Guire units. These are very clearly stated by the writer in Ref. A.I.
CINS 286 to 291.
As a consequence the reviewers proposals do not conform to the re-quired protections and especially against the single failure closed of the RCPS valve in the common suction line of the RHR system and are therefore invalid for these circumstances.
They are also invalid for the case of related water solid operation.
Because of non response to his concerns the writer establishes the validity of his proposals.
Resolution:
Reference the comments under " Issues" One concern requiring particular response is CIN 289 allowing all pumps to be de-energized for at least one hour.
The reviewers, only response is that this will be limited to once every 8 hrs.
In response to the writer's request for a definition of the related circumstances and an analyses of how the plant would respond to transients and accidents under these circumstances to ensure accept-able levels of protection and thereby safety, no response has been provided, This absence of a response means that the plant would be placed in an unana-lyzed safety condition and is thereby Unacceptable.
A fully protected safety status must be established to meet the safety requirements of this need but without the analyses requested this is no possible and thereby the proposed plant status is Unacceptable.
Furthermore, proposed restart of the RCP'S under the above circumstances re-quires evaluation and TS constraints to ensure acceptable levels of potential return to limited nuclear power levels causing overpressurization of the RCS.
Again, the Reviewers' comments on non-testing of alarms and flow rates and other parameters necessary to establish protective system performance, violates the 10CFR36 requirements which are that the LCOS' represent the lowest functional capability or performance levels of equipment required for a safe operation of the facility, and that Surveillance assures that these are met.
Surveillance is not intended merely to ensure that the equipment is functioning only, but also that it is capable of operating to LCO performance requirements, and this is absent from many important surveillance requirements on equipment for which performance can be significantly impacted between In Service Testing (IST) pe-riods:
Also the writer finds no test programs in STS Chapter 5.
Furthermore, redundant alarms are a necessary safety related warning element enabling evalu-ation of timely manual protective action to limit the consequences of the event to acceptable levels within the licensing bases; otherwise the regulations re-quire as a first priority the use of automatic responses for all Occurrences other than Accidents:
3-19 Robert B.A. Licciardo December 1990
Detailed Comments by R.B. A. Licciardo on T Murley Closure of OP0 Issues Regarding the Mc Guire Technical Specifications, dated September 10, 1990.
The Reviewers' comments on available thermal capacities, and alternative manual actions and equipment required to of fset loss of RHR cooling, are very general-ized statements unsupported by analyses and TS for the related equipment, and are therefore speculative and thereby invalid and Unacceptable as a licensing basis for protection against the single f ailure of the RCPB isolation valve which is the licensing bases for Mc Guire.
A fundamental cause for the general problems being experienced by the industry in modes 4-6 is the seriously f ault-ed perceived unimportance of these apparently benign circumstances by the Re-viewerb, and historically is the reason why the Midloop cooling events at Diablo Canyon and the loss of cooling at Vogtle occurred in completely unpre-pared circumstances, after the writer had addressed both situations during 1983 and 1984 in his Mc Guire Review and were rejected by the NRC as being unimportant.
3-20 Robert B. A. Licciardo December 1990
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o a
Detailed Comments by R.B.A.- Licciardo on T. Murley closure of DP0 Issues Regarding the Mc Guire Technical Specifications, dated September 10, 1990.
CONCERN 20B, TS SECTION 3/4.7.4, STANDBY NUCLEAR SERVICE WATER POND (ULTIMATE HEAT SINK)
The ultimate heat sink is essential for final heat rejection to meet Regulatory requirements, including the requirement to be able to cool the plant down to cold shutdown conditions, and subsequent refuelling; and this necessitates special LC0 requirements to ensure continuing operability which may not be im-mediately apparent to those unfamiliar with their conceptual and detailed de-sign, and operating characteristics during the course of a cooldown, and especially to and in Modes 5 and 6.
Furthermore, the Ultimate Heat Sink may take many different forms.
From this experience, the Writer rejects the propo-sition that TS for other dependent systems will ensure satisfactory operation for the Ultimate Heat _ Sink, when-the multiple critical LC05 and related sur-veillance necessary to ensure Acceptable performance are absent together with the necessary definition and Authorities to ensure that they are met to safe-guard the integrity of the fuel in a fully controlled environment under these circumstances.
It must be recognized, that the remaining single Ultimate Heat Sink may only_be a single pool which has been designed at minimum cost and thereby minimum ther-mal storage capability and that after a cool down to Mode 4 and then into Modes 5 and 6, the many operating design limits are being encountered. Or it may con-sist of a cooling tower with related cooling tower pond and many active compo-nents, again operating at their design limits.
Furthermore, that every pro-tective i:ynem in the plant remains dependent on the operability of that single heat sink.
The importance of this system is dominant and Regulatory requirements necessarily place it in the TS.
4 3-21 Robert B.A. Licciardo December 1990
Detailed Comments by R.B.A. Licciardo on T. Murley Closure of DP0 Issues Regarding the Mc Guire Technical Specifications, dated September 10, 1990.
CONCERN 218, TS PAGE 3/4.9-11.
REFUELLING OPERATIONS-LOW WATER LEVEL Comments:
The Reviewers do not respond to the detailed deficiencies of the TS vis a vis Regulatory Requirements, as required for a valid evaluation.
Their comments are absent the required Regulatory and related Technical analyses and are therefore speculative and without merit.
Furthermore the Reviewers have not evaluated writer's Concerns of his OP0 re-view to Ref. A.1 on pages 107 108 and 109 and related CINS 399 through 405, documented and formulated from within the licensing bases for the Mc Guire units; they must therefore remain valid.
Since little or no effective change is apparently proposed for the NSTS, the proposed NSTS will be seriously deficient.
And because of their utmost impor-tance in protecting what has been a significant set of events occurring under this Mode 6, and related Mode 5, represents a serious invalid deficiency by the NRC in the necessary exercise of their responsibilities to Public Health And Safety.
3-22 Robert B.A. Licc'ardo December 1990
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Detailed Comments by R.B. A.- Licciardo on T. Murley Closure of DP0 1ssues.Regarding the Mc Guire Technical Specifications, dated September 10, 1990.
CONCERN 29A, TS PAGE 3/4.7-4:
AUXILIARY FEEDWATER SYSTEM Comments:
This CONCERN is-also identified as CINS' 364-368 and 369 in Ref.A.I.
Since the R6 viewer's have chosen te not allow for the necessary Mc Guire licensing Basis Protection against Transients & Accidents in Mode 4, but refer to the ' dominant deficiency of the existing TS of providing only for potential loss of Decay Heat Coo' ling in this Mode, their review is incomplete and invalid.
As an example of the reviewers' deficiencies, the licensing basis for~ opera-bility of the steam driven -auxiliary fead water pumps-in Mode 5 -is provided under CIN 365.
l_
Concerning the Steam Line Pressure Low signal, the pressure drop across the l-nozzle is the largest and most significant with the double ended steam line break, not the least as proposed by the reviewers.
This therefore does result in earlier Protective actions than.if the pressure taps were taken from up-stream of the Nozzles, and together-with the 7 sec.
closure of the Main Steam line. Isolation valve ensure isolation of the remaining three SGs within approx.
10 secs.
Available information would indicate ultimate blowdown times of up to 25 sec.
for the ruptured SG, and not-a few sets.
The writer's question was i
directed to establishing the residual pressures in the remaining three steam generators to verify that since the TOAFWP,_ would by design be finally operat-ing from one of these units, would the related residual steam line pressures throughout_the event ensure AFW flows consistent with related licensing bases J
l analysis assumptions and especially the since the TS.
LCO.
for the pump speci-fies Operability at a pressure.of greater than 900 psig.
Reference CIN 365.
This information requested by the Writer was not provided by the Reviewers, but research into related-Topical Reports by the Writer shows resulting SG pressures
-of approximately 780 psia compared with the 900 psig required for surveillance testing in the TS.
Furthermore lesser operating values down too 125 psi are required at_the bottom of Mode 3 which is a normal operating requirement, and 4
-in modes 4 and 5 lesser values will be the operating environment in the event
- of-the single failure closed cf the _RCPB valve and in the event all power is
. lost-ie., a Station Blackout (5B0).
Therefore the writer's particular concerns in this area have not been covered by the reviewers and the need for TS changes have been confirmed Furthermore all these concerns relate directly to other TS issues under CINS 117 through 124 to which the same comments thereby apply.
3-23 Robert B.A. Licciardo l-December 1990 l
Detailed Comments by R.B. A. Licciardo on T. Hurley C?csure of DP0 Issues Regarding the Mc Guire Technical Specitications, dated September 10, 1990.
CONCERN 30A, TS PAGE 3.4.7-8:
MAIN STEAM ISOLATION VALVES Comments; The licensing bases for Main Steam Line Break in Mode 4 does require Protective Actions to terminate the event, and for the following reasons:
1.
The earlier propositions of the Reviewers to Concern ISB, Question 8a have been shown to be invalid so that protective actions must ce teken to limit consequences to Acceptable values and this requires the atlated MSIV isolations in this Mode 4, 2.
Even though safety analyses were to show less severe consectorces under Mode 4 with related protective actions as proposed by the Leviewers, the protective actions must be available to ensure this lesser value.
Further-more, without the protective action all SGs would blow down in an-unana-lyzed unprotected event which is unacceptable.
And the consequences of this could be disastrous in many respects by the now ootential inf ringe-l ment of multiple Safety Criteria, and especially if the resulting break i
and blowdown was to outside containment, j
l j
i l
1 3-24 Robert B.A. Licciardo December 1990 l
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e-Detailed Comments by R.B. A. Licciardo on T-. Murley Closure of OP0 Issues Regarding the Mc Guire Technical Specifications, dated September 10, 1990.
CONCERN 31A, PAGE 3/4.7-8a.
STEAM GENERATOR POWER OPERATED RELIEF VALVES (SGPORV)
Comments:
A careful read of the CIN 376 will show that the reason for.the reactor power level of 20% in natural circulation is that although the permissive P-7 set point for reactor trip on loss of all RCPS is set at 10% nuclear power, there is a verified maximum error.of an additional 10 percentage points in the related instrument channels giving a necessarily conservative evaluation at 20% to be used for reactor nuclear power in any safety evaluation.
There is a faulted interpretation by the Reviewers on the representation of Atmospheric Dump Valves (ADV) and the SGPORV.
In his OP0 review both names have been applied by the writer to the same set of valves which are installed downstream of the Steam Generator Safety Valves (SGVS) (but upstream of the Main I
Steam Isolation Valves) (MSIV) and which are steam generator power operated
-relief valves (SGPORV) with a relieving capacity of 10% steam flow and as de-scribed by the Reviewers and the writer earlier in this review and during nor-mal operation are set to actuate during normal operating transients to minimize or prevent the opening of the first SGSV.
The SGPORVS are safety related and thereby required to be included in the TS as described earlier under CONCERN 18A, with the alternate title of ADVS, The confusion arises over the presence of an additional system of Dump Valves which are non-safety related and which are-located downstream of the MSIV's.
A principal component of this system is a dump valve capacity of 10% which exhausts to the turbine condenser, prevent-ing unnecessary loss of steam from the system.
This dump capacity is the mini-mum required to control the plants' heat release during startup, cooldown, hot standby, hot shutdown, and physics testing of the-reactor during normal reactor operations.
However on loss of offsite power and or the condenser this system cannot be used and thereby SGPORV'S have been provided and are necessary to enable the normal safe shut down to the Regulatory Requirement of Cold Shut Down (Mode 5).
Thereby the Non-Safety Related (Non Atmospheric) Oump Valves (NSROV).of 10% capacity are not included in any licensing bases safety analyses and therefore have no place in the TSs.
And the safety related SGPORVS which are easily confused with these and for which-the description Atmospheric Dump Vahe (ADV) has been used are required to be in the NSTS, The resulting con-fusion should be eliminated by usirg the-term Steam Generator PORV's in the-NST5.
3-25 Robert B,A. Licciardo December 1990 i.
+
i Detailed Comments by R.B. A. Licciardo on T. Murley Closure of DP0 Issues Regarding the Mc Guire Technical Specifications, dated September 10, 1990.
CONCERN 32A, TS SECTION 3/4.7.3..
COMPONENT COOLING WATER SYSTEM CIN 378 fully describes the Regulatory Bases for this system including specific operability and operating requirements in Modes 5 and 6.
The system's primary importance for inclusion in the TS can be measured by it being specifically required in the Regulationi, and following that, the Dominant importance in ensuring the operebility cf every protective system and related set of elements on its particula, redundent and independent train.
Further, the writer finds it completely Unacceptable that Reviewers assigned to this task propose that two operable component cooling water systems are not required by this licensee in modes 5 and 6 as the two systems are operating in an interconnected manner so that only one set of pumps are needed.
THIS IS IN COMPLETE VIOLATION OF REGULATORY REQUIREMENTS AND THE RELATED PROCEDURE SHOULO BE IMMEDIATELY WITHORAWN AND REPLACED WITH TS ASSURING NON-CONTINUANCE OF THE PRACTICE.
THE REVIEWERS' STATEMENTS JUSTIFYING THIS HAVE NO SUPPORTABLE EVALUA-TION OF LICENSING BASES QUALITY AND REFLECT AN UNACCEPTABLE UNDERSTANDING 0F NUCLEAR ENGINEERING AND ITS RELATIONSHIP TO REGULATORY REQUIREMENTS FOR NECES-SARY AND ACCEPTABLE PROTECTION AGAINST LICENSING BASES TRANSIENTS AND ACCIDENTS.
Finally, TS are not written in an invalid manner to facilitate maintenance and system modification.
On a nuclear facility arrangements for maintenance and inspection are to be designed around the special features of licensing bases requirements for Nuclear Power Plants with its very special focus on Public Health And Safety as the first concern.
And in this respect the proposal by the Reviewers is faulted and thereby Invalid.
If the licensee wishes to propose special circumstances then he is must propose compensating factors which will provide the same level of protection as provided by the licensing bases and this would be subject to review and approval if acceptable.
But without such evaluation and related change to the TS this would be invalid and thereby a violation of the Licensing Bases, i
3-26 Robert B.A. Licciardo December 1990
Detailed Comments by R.B. A. Licciardt on T. Murley Closure of DP0 Issues Regarding the Mc Guire Technical Specifications, dated September 10, 1990.
CONCERN 33A, TS SECTION 3/4.7.4 SERVICE WATER SYSTEM.
The comments for this system are exactly the same as those for the previous CONCERN 32A with the description of the system replaced by " service water i
system".
i 3-27 Robert B.A. Licciardo December 1990
e a
Detailed Comments by R.S.A. Licciardo on T. Murley Closure of DP0 Issues Regarding the Mc Guire Technical Specification:,
dated September 10, 1990.
CONCERN 35A, TS 3/4.9.8:
RESIDUAL HEAT REMOVAL AND COOLANT CIRCULATION--
HIGH WATER LEVEL The licensing bases for Mc Guire, and any other facility, requires protection against a single failure in the RHR system with a related safety evaluation and establishment of necessary protective actions to limit the consequences of events to Acceptable levels.
In their response the reviewers have provided none of these requirements and is thereby unacceptable.
Thereby the writer's documented safety evaluation under Ref. A.1 TS 3.4.9.8 including related CINS 391 to 397 remain the only valid bases for related Regulatory Actions including related TS.
The speculative proposals by the reviewers have no validity in Nu-clear Reactor Regulation and are thereby Unacceptable.
1 l
i 3-28 Robert B.A. Licciardo December 1990
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Detailed Comments by R.B.A. Licciardo on T. Murley Closure of DP0 Issues Regarding the Mc Guire Technical Specifications, dated September 10, 1990.
CONCERN 36A, TS PAGE 3/4.9--11:
REFUELLING OPERATIONS --LOW WATER LEVEL Reference the writer's comments under CONCERN 35A; they remain unchanged for this CONCERN except for the Subject.
Thereby the writer's documented safety evaluation under Ref A.1. TS PAGE 3.4.9--11 including related CINS 399 to 405 remain the only valid bases for related Regulatory Actions including related TS.
The speculative proposals bj the reviewers have no validity in Nuclear Re-actor Regulation and are thereby Unacceptable.
With respect to the Vogtle situation which occurred under these conditions, the staff has stated that none of the writer's issues addressed Station Blackout (SBO) so that effectively there was no original contribution froni this set of concerns by the writer.
Again the staf f has completely missed the central issues of any regulatory requirement to ensure that the facility WILL NOT BE PLACED IN A COMPLETE LOSS OF SAFETY RELATED POWER FROM THE WORST CASE SINGLE FAILURE, AND OURING A SEISMIC EVENT, or any other of numerous uncontrollable of f site events which could have resulted in the same loss of of f site power that occurred.
THE TS's APPROVED FOR V0GTLE WERE FAULTED IN ALLOWING THE SB0 TO GCCUR AND WITHOUT ANY EVALUATION OF THE CONSEQUENCES, If the NRC staff had accepted the Writer's Concerns and related Safety Analyses when originally issued in 1984, and applied, the SB0 would not have occurred.
These issues of the Votgle event together with those of the Diablo Canyon event were reported by the writer in this TS review of 1984 and evaluated by the staf f as being unimportant and rejected for any further consideration.
After the DP0 they were given the lowest priority, and this continued after the Diablo Canyon event until James Sniezek and later Dr. Murley accelerated the review to its current status in early 1990, after the Braidwood LOCA in Mode 4 event (also principally considered in the McGuire TS Review), and at the same time initiated the now major research program in the area of reactor risk in these Modes.
The reluctance of reviewers to treat this current assignment in a complete and regulatory manner and the manifest unwillingness to accept the writer's earlier work for review in 19B4, and later for the particular case of the Diablo Canyon event, and now for the Votgle event as well as the Diablo Canyon event, would make the NRC staff potentially culpable of serious deficiencies in the performance of their primary responsibility for Public Health and Safety.
D e er I
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Detailed Comments by R.B. A. Licciardo on T. Murley Closure of OP0 Issues Regarding the Mc Guire Technical Specifications, dated September 10, 1990.
CONCERN 38A, TABLE 2.2-1:
REACTOR TRIP INSTRUMENTATION SETPOINT/ POWER REACTOR TRIPS BLOCK,P-7.
The writer finds the requested clarification of this item (CIN-34) inside the Bases is acceptable.
l i
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t l
l 3-30 Robert B.A. Licciardo December 1990 l
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Detailed ' Comments by R.B. A. Licciardo on T. Murley Closure of DP0 Issues Regarding the Mc Guire Technical Specifications, dated September 10, 1990.
CONCERN 3B, TABLE 2.2-1:
REACTOR TRIP INSTRUMENTATION SETPOINTS-LOFS OF PROTECTION USING LOW POWER BLOCKS CommentsonIssue(s)
This OP0 concerns the total problem of the effect of the P-7 permissive block-ing a number of protective reactor trips at low power including zero power in Modes 2,3,4 and 5, and the potential adverse consequences which have not been evaluated except in Mode 2 alone when the power levels addressed are invalid.
Comments on Resolution:
Under CINS 32 and 33 the writer shows that the only available licensing basis analyses for natural circulation is at a power level of 5%, and not the 10%
-quoted by the reviewers.
Furt' ur, under CONCERN 31A, we have confirmed a con-servative power for safety analyses under these conditions of 20% Rated power, making speculative the proposition by the reviewers of acceptable responses to T&A's under these circumstances.
The writer has described a large number of' circumstances under which unsatis-factory: responses to T&A's can occur, but except for the case of the pressur-izer water level trip these have not been addressed.
In fact a number of Events for Assessment have occurred since the writer's DP0, which relate di-I rectly to the effects of P-7 in blocking these trips.
One of these events was the Tripping of all RCP'S "below the P-7 set point" and which resulted in an unexpected power'and pressure surge for the reactor.
For the pressurizer water level trip, the reviewers have not recognized that a primary protective action is that of overpressure protection of the RCS, and that it is not blocked by the P-7 permissive, whilst the high water level trip wh_ich is a back up for that protective action is blocked.
The_ writer did address the question of the automatic water level controller for the pressurizer and showed that for failure of 2 channels of this non safety related system below the P-7 set point, the pressurizer level would reach the trip point in 1/2 hour whilst the surveillance of'the reading is once a. shift so that the reactor is not adequately protected by this manual action in the i
absence.of the trip.
Furthermore, the writer's ~ propositions for its substan-
_tive benefits as an automatic reactor _ trip for T&A's below the P-11 setpoint before water solid operation, haves not been addressed by the Reviewers.
In their comments the Reviewers have not addressed most of the significant safety concerns of the writer-in relation to loss of_ reactor trips from the presence of the relatively " low power" blocks, namely P-7, and P-8, which are presented and evaluated by the writer under CINS 32, and 36-40 of Ref. A.I.
Substantive related materials are also discucsed under CINS 80-88.
Under these conditions, the writer evaluates his concerns as valid and requiring _
-action, l
3-31 Robert B.A. Licciardo December 1990
O 4
Detailed Comments by R.B.A. Licciardo on T. Murley Closure of OP0 Issues Regarding the Mc Guire Technical Specifications, dated September 10, 1990.
QUESTION SB, CONCERN 12B, TABLE 3.3-3:
ENGINEERED SAFETY FEATURES ACTUATION SYSTEM (ESFAS) INSTRUMENTATION The Reviewers' action is Acceptable.
3-32 Robert B. A. Licciardo December 1990
O 4
Detailed Comments by R.B. A. Licciardo on T. Murley Closure of OP0 Issues Regarding the Mc Guire Technical Specifications, dated September 10, 1990.
CONCERN 10A, TS PAGE 3/4.3-ITEM 6c:
SOURCE RANGE NEUTRON FLUX Comments on Resolution:
These Items can also be identified as CINS 77,78 and 79.
Item 1:
The response required by the writer is:
For Mc Guire, the source range and intermediate range-high neutron flux trips, and related trip systems are not Qualified as Safety Related, and the only Safety Related power level trip system available to protect against return to power T&A's in Modes 3-5, when the reactor trip breakers are closed, is the Power Range Neutron Flux Trip-LOW Set Point which is used to protect against these same events in Mode 2:
And therefore TS are required for operability of this Power Range Trip in these Modes 3-5 for these same T&A's.
Furthermore, in answer to the writer's ques-tions concerning the FSAR requirement for the Source Range and Intermediate Range Neutron Flux Monitors and related trips to also be in the TS under the same circumstances, they can serve as a diverse trip system, although they are not Qualified as safety related, under Non-Accident Occurrences when Environ-mental ef fects will have no ef fect on their operation.
Item 2:
What has not been highlighted by the reviewers is that when the con-trol rod system is incapable of withdrawal, the source range monitors are re-quired with their alarm systems to be operable in these Modes, and also Mode 6, to protect against the Boron Dilution Event which has been discussed in this Review under Plant Specific Question 3, TS Table 3.3.1 Item 6, or CINS 74, 75, and 76.
3-33 Robert B. A. Licciardo December 1990
e Detailed Comments by R.B. A. Licciardo on T. Murley Closure of DP0 Issues Regarding the Mc Guire Technical Specifications, dated September 10, 1990.
CONCERN 100, TS PAGE 3/4.3-2:
P-11 INTERLOCK--NEGATIVE STEAM LINE PRESSURE RATE-HIGH SIGNAL This item is also identified as CIN 87 and is related to CIN 104.
The issue is broader than as summarized, as it also includes evaluation of the alternate modes of initiating the reactor trip from the Containment Pressure-High Signal both on small and large line MSLB breaks both inside and outside containment.
The response by the reviewers is invalid as the Steam Line Pressure Negative Rate Signal does not initiate reactor trip.
Furthermore it does not initiate Safety injection for reactivity control of the event and also does not initiate containment isolation; this signal isolates the Main Steam Line Isolation valves only.
Under these circumstances, this Question focussed on the minimum size break which would not initiate the containment high pressure signal and all the related protective a ions and therefore be absent automatic protec-tion, and especially when the break occurs outside of containment.
The response is f aulted and thereby invalid and unless other protective actions considered else where in the writer's review were adopted, would leave the plant unprotected.
3-34 Robert B.A. Licciardo December 1990
. o-Detailed Comments by R.B. A. -_Licciardo on T. Murley Closure of DPO-Issues Regarding the' Mc Guire Technical Specifications, dated September 10, 1990.
CONCERN 14A, TABLE 3.3-3:- ESFAS INSTRUMENTATION.
CONTAINMENT ~HIGH-HIGH SIGNAL IN MODE 4 This item is also identified as CIN 107.
Without separate evaluation by calculation, or a related reference, the writer does not accept the proposition that there is insufficient energy release on a MSLB_or LOCA.in Mode 4 to increase pressure inside containment to the Contain =
ment = High-High-Setpoint of 2.9 psig for Mc Guire Units, and especially when the maximum pressure inside containment is calculated at 15 psig for accident conditions. As one of the principal contributors to total energy insideLthe containment, the energy per pound of saturated water at-425 psig is approx.
80% of that in the SGS at 1050 psig and approx. 60% of that in the RCS at 2235 psig.
And in the case of. the MSLB with the current TS, there is still a' return to nuclear power providing additional energy.
Further it is this signal which-initiates Containment Spray in addition to Deck Recirculation Fans, and not the containment high signal as stated by the reviewers, and so the pressure sup-pression available from its' operation will not become available until this t
set point'is reached, l
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L
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d 3-35 Robert B. A. Liccisedo December 1990
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Detailed Comments by R.B. A. Licciardo on T. Murley closure of OP0 1ssues Regarding the Mc Guire Technical Specifications, dated September 10, 1990.
CONCERN ISA, TABLE 3.3-4:
ESFAS INSTRUMENTATION SETPOINTS-INCLUSION OF NEW ESF FUNCTIONAL UNIT IN THE TS.
The comment by the reviewers is incorrect as the writer has provided the rea-sons for the proposal under CIN 164.
Whereas the current TS provides for only one Functional Unit, Feedwater Isola-tion, there are in fact two elements to this activity, namely Trip of all Feed-water Pumps and Main Feedwater isolation.
Further whereas the trip of all Feedwater pumps is initiated by only two sets of Protective Logic, that of Main Feedwater Isolation is initiated by four logic sets.
So that the distinction need to be made and the related logic systems subject to the appropriated TS LC0 and Surveillance Requirements feedwater system and by four separate sets of ESFAS logic.
3-36 Robert B.A. Licciardo December 1990
e
' TABLE 3.1 LIST OF MINIMAL SET OF GENERIC ACTIVITIES AND ACTIONS ARISlNG-FROM VARIOUS ENTITIES, AND DERIVING FROM THE R.B.A. LICCIARDO MC.GUlRE TS-REVIEW OF 1984 (REF A.1).
( A SUBSET OF TABLE 1)
A cc o r a ti J TEMNO GENERIC GENE'RICW5 GENGT UDY f 4EW5T5 WSTL ASHT AD 01HAETil 1M510
.t 2
N515+
TMF -
?
2 N515+
TMF o
6 N5T5+
R5BWT5 TMF
.7.
7 f l5T 5+
R5DW15 TMF D
G-N5T5+
RSBWTS:
TMF 20 21 A
R50WT5 TM To 27 G5WN 27 2 E' G5WN R E DN T 'S 31 32 GW G5 RSBWT5 32 35 GW G5 RBDW15 37 34 G.A GW.A IGTS
.A R5BWT5 TMF T4 35 G.A GW.A
.A R5bWT5 35 36 NSTS
.A 36 37 N5T 54 A
TMF 37 30A G.- A GW.A GNU IGT$+
.A TMF 39 35U b.i GW.A GSN NETS'
.n TMF 41 41 G.R5b N5T5+
TMF 4?
42 N5154 TMF 43 4"
tJ5T54 TMF 44 44 N515+
TMF 45 45 G
N5T5+
A REBWT5 TMF 51 51 G.RSP S2 52 N5T5+
TMF 53 50 NT5+
TMF SS 55 N5T5+
TMF b6 b6 N5T5+
TMF 5'I 57 N5T5+
TMF
(
58 55 IGT5+
TMF T.9 59 N 515 +
TMF 60 60 N515+
TMF 61 61 N5T5+
TMF 6?- 62 N5T5+
A TMF i
e3 63 Il5T5+
64 64-N5T5+
e5 65 1615+
W.W15 TMF 67 e7 G.A GW.A 14S15+
.A RSBWT5 TMF 68 6 E.
G.A N5T5+
.A.
R5LtWT 5 TMF 69 ~69 G.A GW.A N5T5+
.A RSBWT5 TMF 70 70 G.A GW.A N5T5+
.A R5BWT5
'TMF 71 71
.G.RSS NSTS+
R5DWT5 TMF 72 72 G.A GW.A NSTS
.A R5bWT5 TMF l
73 77 G.R5B N5T5+
R5BWT5 TMF l
74 74 N5TS+
4 TM l
75 75 N5TS A
TMF 77
-77 G.A GW.A.
.A TNF 18 78 G.A GM.A f l5T 5 +
.. A TMF l.
79 79 G.9 GW.A
.N5T5+
.A 01 83 R5EWTE D3 61 G
GW N5T5 TMF B i!
G4 N5TS TMF=
GS US G.R5D l
Oc-86 G.RLD TABLE 3.1 l
l
m e
TABLE 3.1 LIST OF MINIMAL SET OF GENERIC ACTIVITIES AND ACTIONS ARISING FROM VARIOUS ENTITIES, AND DERIVING FROM THE R.B.A. LICCIARDO MC.GUIRE TS REVIEW 0F 1984 (REF. A.1).
(A SUBSET OF TABLE 1)
- st e o r ci n ITEMNO GENERIC GENERICWE GENSTUDY NEW5T5 WSTS ASHT AD OTHACTN TMSIO l 07 97 G.A GW.A N5T5+
.A R5bW15 I
BE 88 G.A G5W.A N5T5+
.A W.RSBiaTS 9j 91 0.RGB c,' 2 92 G.W A
F4BWT5 TM 94 94 G.R5E 96 96 G.RSD A
TM 97 97 G.RSD TM
)
90 98 G. RED A
TM 99 984 G.R5B A
TM 105 103 G
106 104 G
W D C, N5T5 R5BWT5 TMF
.108 106 G.R5B N5TS TMF j
109 107 N5154 TMF
- 110 108 G.A GW.A
.A 112 110 N515 TMF i-113 111 N5T5 F E PW15 TMP Ajo 112 G.R50 N5TS TMF 119 117 G
GW NST5+
TMF 120 116 G
GW N5154 1MF 121 119 G
GW N3T5+
TMF l
122 120 G
G k' N5154 TMF 123 121 G
GW MST5+
TMF 124 122 Cs GW N5T54 RSBWT5 TMF 125 123 G
GW N5T5+
TMF J26 124 G.RSB 12G 12e G.RSB 129 127 N5TS A
TM 130 I?E N5T5 4
TM 131 129 G.A GW.A N5TS
.A TMF J34 1 3 '.
05 135 133 G5 136 13e G5 140 13D G.A GW.A N5T5+
.A 141 139 N5T5+
147 l o t.
N5T54 A
TM 148 146 N5T54 A
TM 150 140 N5TS TMF 165 15.
i.,
GW N5T5+
ITEM 122 150 157 G
A TM 160 159 G
l 162 le0 G. RED 166 164 G.A-GW A N5T5+
.A TMF 167 165 H5T5+
TMF 169 16e N5TF+
TMF 169 167 NET 5+
TMF 170 168 N5T5+
TMF 171 169 N5T5+
TMF I
JT2 j'O NETL+
TMF 17
- 71 N5T%+
TMC J 31
~2 C.A GW.?
C,L N515 +
.A N
TMF 272
.:2 0 G.A GW.A GL NET 5+
4 it_
TMF l
2 3
231 OA GW.h GL N515 i-
.A Wt TMP TABLE 3.1 (cont) i
L TABLE 3.1 LIST OF MINIMAL SET OF GENERIC ACTIVITIES-AND ACTIONS ARISIN3 FROM VARIOUS ENTITIES, AND DERIVING FROM THE R.B.A. LICCIARDO MC.GUIRE -TS REVIEW OF 1984 (REF. A.1).
(A SUBSET OF TABLE 1)
Rec o r d 4!
1TEMND GENERIC GENERICWE GENETUDY NFW5TS W51 E A%1 M m H42TN TMSID 234 23.c G.A GW.4 GL N5T59
,A W
TMF 23b 237 G.A GW.A GL NET 5+
.A Wu TMF 2.6 234 G.A GW.A GL NET 54
.A W.
TMF
.237 236 G.A GW.A GL NST54
.A' WL TMF 238 236-G.A' GW.A GL N5T5t
.A WL TMF 239 237 G.A GW.A GL N5T5+
. (-
WL TMF 240 23G G.A GW.A GL Nb7 54
.A WL 1MF r-241 239 G.A GW.A GL N5T54
.A uk 1MF 242 240 6..A GW..
GL N5T 54
..A W,.
7MF 243 241 G..A GW..A GL N5T5+
..A Wi.
TMF 244 242 G..A GW..A GL NETbi
..A WL,R5BWTS TMF' 24S 243 G..A GW. A G!,
NET 54
.4 WL R5EWT5 TMF 24e 244 G..A GW..A GL N575+
..A WL TMF
'i 247 24S G..A GW
,A GL NET 54
..A WL TMF 240 246 G..A CW..A GL N E.1 6.
..A WPmv TMF 249
?C G..A GiJ..A GL NETS +
..A FEcsit TMF 250 W
b..A bW..h GL NM ia
.4 6.
1MF 2S1 2 c. 9 G..A-GW..A GL NR s+
..A WL TMF 252 250 G..A GW..A L%
N5T54
..A Wi.
1MF Ob7 251 G..A GW..A GL N5T54
..A UL TMF 254 2N' G..A GW..A GL NBT54
..A in TMF 255 253 G..A GW..A GL NST 54
..A WL TMF
'5e 254 b..A bW..A 6L N5154
..A Wa TMF 2b7 :2b5 G..A GW
.A GL N5T5+
..A WL TMF 205
- 2bt, G..A GW..A GL N5154
..A WL TMF 259 257 G..A GW..A GL N5T5+
,.A t!L TMF 260 25D G..A GW..A GL NGT5+
..A WL TMF 261. 259 G..A
'GU..A GL UST5+
..A WL TMF 262 260 c..A GW..A GL N5TS+
..A WL TMF 263 261 G..A OW..A GL NST5+
..A ik TMF 264 26:
G..A DW..A GL NET 5+
..A WL F.5BW15 TMF 269 267 GL N5T5+
WL 270 20D GL N5TE+
Wm 271 269 UL N5TS WL 272 270 Gl.
NB154 W!. FiiOWTS i
l-l 273 -27i GL-N5T5+
Wu. e,SBWTS -
275 T73 6
GL N5TE A
WL TM 276 274' G
-GL N A T c, A:
WL TM
'277 275
-WDG NST5+'
TMF L2 ? O 276-WOG N5T5+
TMF T7' 277 G.A GW.h WOG N5T54
.4 RSBWT5
-TMF
~260 270 G..A GW..A WOG N5T5
..A RSBWT5
=TMF 201 279 G..A GW..A WOG N 515+
..A TMF 282 2Gs W3G l
2D3 201
~G.A GW.A WOG N5TS+
.A TMF l-284 20i G.A GW.A WDG N5T5+
- A
~TMF 285 CEG
.G.A GW.A WDD N5T5+
.A TMF 286 204 G,A GW.A WOG N5T5
.4 TMF 1
M 2Lb G.A GW.H-WD2 NM 5"
.A TMF 20 28L N5T5+
TMF-267 E
, J 51 E 4 F 60W1 b TMF cw 20.
G.A Gw.-
tGT5+
.A TMF TABLE 3.1 (cont)
y a
TABLE 3.1 LIST OF MINIMAL SET OF GENERIC ACTIVITIES AND ACTIONS ARISING FROM VARIOUS ENTITIES, AND DERIVING FROM THE R.B. A. LICCI ARDO MC.GUIRE TS REVIEW 0F 1984 (REF. A.1).
(A SUBSET OF TABuE 1)
M c o r cis!
3TEMND GENERIC GENERICWE GENETUDi NEW5TS WET 5 ASHTAD 01HALTN TMSIO 2(i j 28; G.A GW.A N5T5+
.A TMF 292 290 C.A GW.A N5T54
.A TMF 297 29.1 G.A GW.A I45T 5+
.A TMF 294 292 G
GW NET 5+
TMF 295 293 G
GW N5T5+
TMF 2%
294 G
GW N5T54 TMF 297 294 G
GW N5T54 TMF 29D 296 6
GW NbT5+
TMF 299 297 G
GW I45T54 TMF 300 290 G
GW N515+
TMF 301 29" N:s T 5 +
TMF i
302 300 N5T54 TMF 303 301 N5T54 TMF 304 302 h5D H5154 TMF 306 30%
N575+
306 304 N5TSt RibilTS TMF 307 COS N5T54 TMF 308 506 RSD G5 NET 5+
R 51 W15 TMF 309-307 G.R5B G5 N5T5+
TMF 310 308 G.RGb G5 N5T54 P5E WT 5 TMF 311 304 G.RSB GE N5T5+
TMF 312 310 G
WOG NET 54 313 311 65 N5T5+
314 312 GS NET 54 31S 3134 G.RSD GE N51b+
316 3131:
G.R5D NET 5+
3J7 314 G5 316 315
-N515*
R51-:WT 5 321 318 0.TM WOG N51b+
A TN 32:
318A G.lh WD5 N5154 s
TM 323 318D G.TM WOG N5T54 A
TM 324 3263 WOO HTT54 325 315D WOG N5T5+
326 318E WOS N5T5+
327-31GF UCG-N515+
328 319 G.TM G5 N515 A
TM 324 320 N5T5+
WST5 W
330 321 N E T 5 +-
W5TS W
331 322 N5T54 W
232 323 G. Rib R5B NET 5+
333 324 G.R5B RED N5T5+
334 325 G.R5D R5b N5T54 335 326 G
c TM
'342 333 REBWT5 342 3F G.RSE 364 3Ss N5T54 TMF 365
~* S 6 NET 5+
TMF 3e_ :
W' NET 5+
TMF T i 35E N5T5 4-TMF Sir W
heli
- TMF 36'
'.. v NET 5+
TMC N5T54 TMF 330 TABLE 3.1. (cont)
,.c TABLE 3.1 LIST OF MINIMAL SET OF GENERIC ACTIVITIES AND ACTIONS Ai'ISlNG FR0i4 VARIOUS ENTITIES, AND DERIVING FROM THE R.B. A. LICCI ARDO MC.GUIRE TS REVIEW OF 1984 (REF. - A.1).
(A SUBSET OF TABLE 1)
% t e c t' ITEMNO EENCPIC GENERIttdE GENSTUDi NE.WST5 WET E AE ri1 AD to t:ACTN TMSIO e
" 1 362 N515 +
TMF
,74 365 G,A GW.A N5T5+
,,A TMF 375 366 G.A GW.A N5T54
..A TMF 770 367 G.A OW,A NET 54
..A 1MF 377 3c9 G.A GW.A NET 5+
..A TMF 376 369 N5Tf+
TMF 384 3 ',7 S G. (*
GW.A LIC'3 NET 5
. />
TMF 385 376 G.A GW.A WCiG NB1L
.A F5EsW15 TMF 386 377 G. (i GW.A l.DG N575+
.A P5EvTE TMF "07 370 G. (i GW.A N5T5+
.A TMF 38E 379 G.A GW.A
.A 359 380 t$15' TMF 390 381 G.A GW.A N5T5+
.A TMF 305 3E9 G. RED 400
-+1 N57 54 TMF-4Of 392 NET 54 TMF 402-
"cc N r,15 4 TMF 40!
394 G..A GW..A NS754
..A TMF 404 395 N D'15 +
TMF 4GS 396 hSTS4 TMF 406 -397
-G..A GW.sA NB15+
..A TMF
-408 399 G.A GW.A NST 5 +.
-.A TMF 409 400 E.A GW.A NL154
.A TMF 410 401 G.A GW.A H5T5+
TMF 411 402 I ET 5" "f MF 412 403 G.PSD NET 54 TMF g -
413 404 N5T5i TMF 4.14 40S G.A Gle. fi N5TS4
.A TMF 415 406 G.RSD NGT5+
TMF 41c
'D7 G.RSP N5tS4 TMF 417 400 N5159 TMF 418 407 H 515
- TMF TABLE 3.1 (cont) i e
x
t TABLE 3.2 LIST OF F,1NIMAL SET OF ACTIONS ON THE EXISTING TS AND WESTINGHOUSE STS ARISlNG FROM VARIOUS ENTITIES, AND DERIVING FROM THE R.B.A. i.lCCIARDO MC.GUIRE TS REVIEW 0F 1984 (REF. A.1).
( SUBSET OF TABLE 1) ft:ercl 11EM50 6ENEF!C SEh!E!Cai SEh5TUI'Y htt515 E115115 kS15 ASHIA! 01hA2th f r.!!(
1 1 h515+ EIS Tr3 22 h5iB+ E15*
Tr:
4 4 ETS 6 6 h515+ E15+
E5hT5 16 7 7 h5T5+ E1!*
E56k15 IMF 26 25 EIS A
TF 31 22 h
65 E15 E51st!
32 33 66 65 E15 15hi!
33 34 6.A 6a.A h515 Elin.A
.A PitiTS iti 34 35 6.A 6h.A E15k.A
.A E5 int 5 37 35A 6.A Gk.A 6W kS15+ Eifs+
.A 16 36 3EE 6.A 6k.A 65m N5T5+ E15s.A*
.A IF 45 45 6
h!15+ E15+
A Ri!*i!
TF 50 50 ETi 62 62 kili+ E15+
A 16 65 e5 h5Ti' ETS t.Lii IP:
t7 67 6.A En.A h515+ Elis.A+
.A F5bl5 1F 15 (!
fA h575+ E156. A+
.A f!bi5 fr3 J
69 64 6.0
!*.A N515+ ET51.A+
.A F5bl5 16 70 70 6.A Ek.A kS15+ E15h A+
.A E5h15 It 72 72 6.A H.A h515 ETh A
.A F5bl5 TF l
74 74 d515+ E15+
A 1r 75 75 h5T5 Ei5+
A 1F 77 77 6.A 6k.A h5TS E15*.A
.A 1F 76 72 6.A Sk.A kET5+ ElBa.A
.A Tr 1
79 79 6.A 6v.A h5ti+ ET5s.A
.A 63 93 6
6k h5fi ET5k 16 1
64 64 h515 Eli IP:
l 67 67 6.A Ba,A A515+ Eib. A
.A Esisi!
65 62 6.A 65=.8 hit 5+ EI5k.A+
.A h.E5f*T!
92 9 6.k ETS A
R5bTE It l
94 h 6.fi!
E15+
A 1"
l 97 57 6.FSF E15+
fr 99 9EA 6.f 5!
EI5+
A IP 106 10b 6.E51' h315 FIS TF 110 105 6.A 6h.A Ei56. A -
.A 111 IM E15k.t!L i
l 114 112 6.R5!
h5TS E15.t>EL TF l
119 117 6
6a h5i5+ ET5:e TF 120 112 6
66 h515+ ET5a+
ir5 121 119 6
6k N5T5+ Ei!
TF 122 120 6
En N515+ ET5+
TF 123 !?!
6 Ek N!15+ ETS 1"
124 122 6
fa h515* EIS Ribi5 IPJ 125 123 6
D h515+ E15+
IF 126 124 6.f52 E15.
125 126 6 E5i EI!
129 127 h515 EIS A
TP 130 12E F515 Ei!
A 1"
131 129 6.A S.A hii5 Elis.A
.A 1"
IF; !!E G.A Ga.A h!TE. ETH.s
.A TABLE 3.2 1
l
TABLE 3.2 LIST OF MINIMAL SET Of ACTIONS ON THE EXISTING TS AND WESTINGHOUSE STS ARISING FROM VARIOUS ENTITIES
- AND DERIVING FROM THE R.B. A. LICCI ARDO MC.GUlRE TS REVIEW OF 1984 (REF. A.1).
(SUBSET OF TABLE 1) ft;cret 11[r4 GihEF.!( !Ch!f1(6I (th!1tti kit!f 5 !!!$115 bil$ A5HA! 01*A:1n Dil:
141 !??
h!16+ (15 147 145 hille !!!t.[Il A
9 lif 141 hili + (Ift.lil A
9
!!4 !!?
(15 11t+ D
!!! it!
I ft hilf+ [156 liif 1;2 162 160 6.555 f1!
let 164 E.8 En.A h!15+ (156.A
.A 9:
174 114 (15+
A It 177 175 Ell
- A lr ID 17f 115
!!! !?9
[11+
A f*
IfB lit (15+
1 It 1H li?
(1!+
A ll* 14e (li 153 141 (1t*
A tt JH 111 (ti+
A tr 200 195 (if la 202 2M (15 28 201 ti!*
A h
2M 2M til A
D 213 211
[15 A
t t.
131 125 6.A 66.4 h
hil!* (15*.A+
.A 6
1,r 132 2M 6.A ts.A 6.
hit!+ [1f 6.4
.A 6
D:
6.A 6t.A h
h!15+ [156.4
.A n.
pr 34 23; 6.A fi A
[.
kilt + (15m.A.
.A t.
1" 235 P 6.A 66.A 6.
Mil + [15k.4
.A 4
pr 23t 234 6.A ls.A h
h!15+ llik.h
.A 6
D*
2:7 !!!
6.A Gb.A
(.
hil!* [156.A+
.A 4.,
1**
!!6 Ut 6.A (n.A h
kill [1b.h
.A 6;
p_r
!!i !!7 6.A (6.A 6.
h!!$* [116.A'
.A 6
itt 140 132 6.8 66.4 h
kill + [156. 4
.A 6
DJ 241 M4 6.A 66.A L
k5if
- lif t.4
.A 4
1Ff
- 42 240 6..A ft..
6; hili
- liik..h
..A 4.
1" 247 l41 6..A b.. A 6.
kil5+ E1!6..b
..A 6
Dr
- 34 24
6..A (6..A k
hili * [15s..A+
..A 6..f!!**! 1Pt
!45 243 6..A (k..A R
hil!* Ili6..b
..A t.Fibli D*
I44 284 6..A D.. A' k
kil!' [1b..A+
..A 6.
1*f 247 24$
6..A 6t..A h
hili * [ lit..b
..A 6;
IPT 242 241 f..A (b..A R
hil!* [156. 4
.A I!!hi!
Ir!
249 287 f..A fa..A 6'
hill
- Elf 6..b
..A Filsli 1 73 it' its 6..A (k..A h
h!15* [15h. 4
..A t.
IrJ
?!1 244 6..A 66..A R
h516 + (15s..b
..A k.
i9t
!!! 250 E..A 6s..A k
N!15+ fih..b
..A t.
1*t
!!3 M1 f..A D..A h
kil5+ [156..b
..A t.
Irf
?!4 2!
6.J
';n.. A h
k515+ (15t..D
..A 6.
It' 25! 253 6..A ft..A R
h51b (1!6. 4
..A t.
1" 2!t ?!:
(,,A b.,A k
hift* (ii6,.A+
..A t.
1tt 2f' 25!
6..'
D.. A L
A51t* fib. 4
..A k.
It:
2ti 25e 6..A Ei..A R
h!1E+ E155..h
..A k.
D1
?!i ?!?
6..A D.. A 6.
h!15+ [155.. b
..A 6
D:
- e0 M5 5..A h.. A h
h515+ fila..D t.
1"
- il ?!;
6..A b.. A h
hi'D Elf t..b
..A 6
Di
- t?
- :3 E..A b..A 6.
hilE+ Elit.8*
..A 6.
1" TABLE 3.2 (cont)
s TABLE 3.2 LIST OF MIN! MAL SET Of ACTIONS ON THE EXISTING TS AND WESilNGHOUSE STS ARISING FROM VARIOUS ENTITIES, AND DERIVING FROM THE R.D. A. LICCIARDO MC.GUlM TS REVIEW OF 1984 (REF. A.1).
(SUBSET OF TABLE 1) ft:tret 11E N Ethtfl! Elktiital (thilutt httill it!!115 tiffst*it 01rt:1h it!M 2:3 2t1
- 6. 8 Es..A E.
h!!! + [116..?*
t.
1er 2tt it; 6..A 6t..A L
hili + lift..'+
..A 6..lil*1E1*r it; 2e7 EL hi!!+ fil.**
6.
270 *ti h
hi15+ (15.9 6.
271 ;t9 k
h!15 115.6 t.
272 270 h
hiil+ E15. D t..liitti 273 ;71 h
k!1!* (1!.o 6..l!!e!!
27! 273 i
k h516 (156(15+
A t.
I' 276 274 6
h hit! (ilstile A
t.
1*
279 277 6.A fe.A b;5 h!!!* (116.A+
.A f!!all 1*:
Il0 17E 6..A 66..A h3 kill (156..A
..A f!!61!
It
!!! 279 6..A 06..A bM kill + lif t..D
..A iPi 2(; 263 b;B fi!n 2(3 261
(.A (t.A b;*
k!!!* 11!t 4+
.A 1P8 264 it*
f.A (6.8 6:0 hiii* [1fe D
.A 1 *.
21! 2i:
6.A (k.A h*i hiti+ [154.. H
.A 1**
26t 2(4
(.A 66.4 O*
hil! [lin A
.A l'T 217 2i!
E.A (W.A tX h511+ ET!n.P
.e it:
240 26E 6.A 6t.A k!!!* ilit.D
.A 1P:
241 269 6.8 (t.A hi154 (15s.O
.A lar 282 240 6.A
$6.A hili + (1!t.4
.A Tri 253 ;41 6.A (n. f, h!16+ [116.A+
. f.
I P,'
244 292 6
(a h51f+ [15s' it; 25! 243 E
En hi16+ (1tu 1**
29t ;44 E
5 h!15* Elio it:
217 294 6
66 h315+ t1SD 1**
24f 246' A
(t hii!* (1!o 1*r 249 251 6
Es h515+ (1!D l':
M 25E E
Ei h!15+ t15 +
1rr
!!! !!0 0
6:3 kilf
- 11!"
21: 311 si w!!!. [11.h+
314 31; Ei h!15* (15.h*
!!! !!!A
- 6. lit Ei hit!* Eli.e
- 1t 01 t
- 6. Fit h511+ lit.D 3:6 !!i 6.!*
EE hiti Elin A
It 3F 1h h!15+ (Si hit:
6 3:0 :21 hitte fit till n
!!! :22 h515. [15+
6 332 ??)
6 fli f!F hil!+ (1!"
333 !*4 E.lil f.51 kiil+ E1!n 134 3:5 f.f!t fit h!15. Ei!"
!!! !;t E
(15 A
TF 34! ;3e (15 8
1?
!!* 150 E15 A
fr 30 ;!!
til A
in
!c2 !!!
E15 1r 3e4
't.
hiti+ til l '.:
3:: :P kiis' til l':
371 3e!
6.s En.A hi15+ (156.h
..A 1*:
- 'S :e:
E.A si.t hili
- ttit.A+
..A 13 rt 70 E.'
ti.
h!'i+ tiit.+
..A W
?' 3c:
5.4 fk.A hil*+ [lik.h
..A 1**
TABLE 3.2 (cont)
3 TABLE 3.2 L!ST OF MINil%L SET OF AC110NS ON THE EXISTING TS AND WESTINGHOUSE STS ARISi.iG FROM VARIOUS ENTITIES, AND DERIVING FROM THE R.B. A.1.lCCI ARDO MC.GUlRE TS REVIEW OF 1984 (REr. A.1).
(SUBSE1 0F TABLE 1) te m ti 11t*0 Ethitit 6twilltet Eth5TL'It siti151111115 til! Atal&!'(1K:16 1'i lt-
!ia 375 6.&
le,&
06 h!15 ilit.!
.A 1*J 1657e
(.s Ea,&
RB hil! lift.&
.4 Filei!
1's 5te 377 6.6 in.&
K5 hil5+ (1!n.A+
E!!st!
It!
3l7 37i 6.5 16.8 h511+ ilin.k+
l'I
!if 3M 6,&
6t.A
[1!s.A
.A 364 ;[]
hS15+ [16+
1" 19(> 361 6.6 66.&
hill * [Ilt.f+
IH 191 ;!:
[15+
A 1r 397 ili
[11*
A IP 403 !44 6..A 66..&
hilf* (1!.s,.(*
..A IFJ 40t 391 6,.&
Es.,&
h!1}+ [156.,4+
IFJ (ft 199 6,6 66.A hil5+ [Ill.A+
1" c
409 400 6.6 66.&
hi1E+ [118.s*
.8 TH 41(> 4: 1 6,A 66.4 k!15+ (156 &+
4:4 4(t 6,&
En.&
Nil!* (1!h.&+
.A 1"
4:1 151A til.
TABLE 3.2 (cont)
W LIST OF REFERENCES FOR COMMENTS BY R. LICCI ARDO 1.
Letter f rom N. B. Tucker (D.P.Co) to H. R. Denton (NRC) dated September 27, 1982 to the subject of "McGuire Nuclear Station."
2.
Memo f rom 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 Regulatory Commission, Final Safety Analysis Report, Volume 7, Duke Power Company, McGuire Nuclear Station, Units I and 2, Rev 45.
6.
V,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 Stat 4on Units 1 and 2, Rev. 45.
9.
Deleted 10.
U.S. Nuclear Regulatory Commission; Office of Nuclear Reactor Regulation;
" Safety Evaluation Report; McGuire Nuclear Station Units 1 and 2, Duke Power Company," NUREG-0422, on Docket Nos. 50-369 and 50-370, March 1, 1978 11.
U.S. Nuclear Regulatory Commission, Of fice of Nuclear Reactor Regulation;
" Safety Evaluation Report, McGuire Nuclear Station Units 1 and 2, Duke Power Company," NUREG-0422, Supp. 1, on Docket Nos. 50-369 and 50-370, May 1978.
12.
U.S. Nuclear Regulatory Commission, Office of Nuclear Reactor Regulation;
" Safety Evaluation Report, McGuire Nuclear Station Units 1 and 2, Duke Power Company," NUREG 0422, Supp. No. 2, on Docket Nos. 50-369 and 50-370, March 1979.
13.
U.S. Nuclear Regulatory Commission, Office of Nuclear Reactor Regulation;
" Safety Evaluation Report, McGuire Nuclear Station Units 1 and 2, Duke Power Company," NUREG-0422, Supp. No. 3 on Docket Hos. 50-369 and 50-370, May 1980.
34.
U.S. Nuclear Regulatory Commission, Office of Nuclear Reactor Regulation; "Saf ety 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.
3 46
', cw 15.
U.S. Nuclear Regulatory Commission, Of fice 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 I and 2, filing amendment No. 71 to its Application for License for the McGuire NL. clear 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 (HRC), dated Oct. 8, 1981 on the subject of McGuire Nuclear Station, Unit I and submitting copies of Peport identified as " Westinghouse Reactor Protection System /
Engineered Safety Features Actuation System Setpoint Methodology, Duke Power Company, McGuire Unit 1," by C. R. Tuley et al. and dated April 1981 published by Westinghouse Electric, Nuclear Energy Systems, PROPRIETARY.
19.
Westinghouse Electric Corporation, PWR Systems Division " Westinghouse 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 (D.P.Co), dated May 17, 1983 on the subject of OL Condition 2.C.(11)g, Anticipatory Reactor Trip (11.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, 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. Parkar (D.P.Co) to H. R. Denton (NRC), dated Septemaer 18, 1980, re:
McGuire Nuclear Station.
Page 13, Response to 3(e).
25.
Duke Power Company McGuire Nuclear Station, Unit 1, Docket No. 50-369, License No. NPF-9 Startup Report, February 15, 1982.
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 of Inadvertent Boron Dilution Events.
- 27. Westinghouse Electric Corporation, Nuclear Energy Systems Topical Report, Overpressure Protection for Westinghouse Pressurized Water Reactors, WCAP-7769. Rev. 1, June 1972.
3-47
4 u
l
- 28. Westinghouse Electric Corporation for the Westinghouse Owners Group on Reactor Coolant System Over pressurization, 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.
30.
Memo for:
Brian W. Sheron (NRC) f rom Robert B. A. Licciardo, dated June 11, 1984 on the subject of Review of McGuire Technical Specifications.
31.
Westinghouse Electric Corporation, PWR Systems Division Topical Report WCAP-9226, Rev. 1, January 1978, NES Proprietary Class 2, Section 2.
32.
Wettinghouse Electric Corporation, Western Nuclear Energy Systems, " Report on the Consequences of a Main Feed Line Rupture" WCAP 9230 Proprietary.
- 33. Westinghouse Electric Corporat4on, Western Nuclear Energy Systems, Setpoint Studies, Duke Power Company, William B. McGuire Nuclear Plant, Unit 1 & 2, May 11, 1978.
34.
Letter to Thomas Novak (NRC) to H. B. Tucker, (DPCo), " Request for Comments on McGuire Technical Specifications Concerns Resulting from Differing Professional Opinion," dated July 9,1985.
35.
Letter from H. B. Tucker (DPCo) to Harold Denton (NRC), "NRC DP0 Concerns on McGuire Technical Specifications," dated June 10, 1986.
36.
Memo from Robert B. A. Lit lardo, to H. R. Denton on
Subject:
McGuire Technical Specifications:
Request for Completion of Post Resolution Phase of Dif fering Professional Opinion of Mr. R. Licciardo, June 3,1985, 37.
Memo from Ashok C. Thadani to Steven A. Varga on the
Subject:
Resolution of Plant-Specific DP01ssues Concerning McGuire Technical Specifications, dated May 14, 1990.
38.
Letter from H. Thompson to R. Bernero, " Disposition of Concerns Raised by R. Licciardo in his DP0 on the McGuire Technical Specifications," dated May 28, 1985.
39.
Letter from A. Thadani to C. E. Rossi and S. A. Varga on the subject:
" Assignment and Schedules for Resolution of McGuire DP0 Technical Specifi-cations," dated January 26, 1990, 40.
Memo from T. M. Murley, Director, NRR, to Robert Licciardo, PMAS, NRC, on the subject: "Cinsure of Outstanding Technical Specification Concerns Deriving f rom R. Licciardo's DP0 Review of the McGuire Technical Specifica-tions." (TACS 55435/55436/67757) dated Sept. 10, 1990.
41.
USNRC; Standard Technical Specifications for Westinghouse Pressurized Water Reactor (Revision Issue Fall 1981); NUREG-0452-REV-4 dated November 1981.
Westinghouse Electric Corporation, PWR Systems Division, Reactor Core Re-42.
sponse to Excessive Secondary Steam Releases by S. D. Hollingsworth and D. C.
Hood.
January 1978.
WCAP-9261. Rev. 1, NES Proprietary (Class 2) 3-48
"^---------%-
A.1 Letter f rom Robert Licciardo to Brian Sheron, " Review of McGuire Technical Specifications," dated June H, 1984.
A.2.
Letter from Thomas Novak to H. B. Tucker, " Request for Comments on McGuire Technical Spccifications Concerns Resulting from Differing Professional Opinion," dated July 9, 1985.
A.3.
Letter f rom H. Thompson to R. Bernero, " Disposition of Concerns Raised by R. Licciardo in his DP0 on the McGuire Technical Specifications," dated May 28, 1985.
A.4 Letter from H. B. Tucker to Harold Denton, "NRC DP0 Concerns on McGuire Technical Specifications," dated June 10, 1986.
A.S.
Memorandum from Thomas Murley to Robert Licciaroo, " December 7,1983 Dif fering Professional Opinion," dated December 29, 1989.
A.6.
WCAP-8745-P-A, " Design Bases for the Thermal Overpower AT and Thermal Overtemperature ai Trip Functions," dated March 1977.
A.7.
NUREG-0964, " Technical Specifications McGuire Nuclear Station Unit Nos. I and 2 " dated March 1983..
A.8.
Letter from William Parker to Harold Denton, " Westinghouse Reactor Pro-tection System / Engineered Safety Features 1,ctuation System Setpoint Methodology, Duke Power Company, McGuire Unit 1," dated October 1981.
A.9.
Duke Power Company, McGuire Nuclear Station Final Safety Analysis Report
- Volumes 5, 6, 7, 9, 10 and 12.
A.10. ANS-56.2, " Containment Isolation Provisions for Fluid Systems," 1976.
A.11. Generic Letter 85-05, " Inadvertent Boron Dilution Events," January 85.
A.12. Letter from George Lear to D. C. Switzer, " Millstone Nuclear Power Station Units 1 and 2,"
dated June 1977.
A.13. Letter f rom E. P. Rahe, Jr., Westinghouse Electric Corporation, to Mr. D. Eisenhut (USNRC) on the Subject of Number of Operating Reactor Coolant Pumps in Mode 3, July 9,1984.
A.14. Memo f rom Robert Licciardo, PSB, NRC, to Steven A. Varga, Director, en the subject: Comments on Resolution 6 and Plant Specific DP0 Issues Concerning McGuire Technical Specifications, dated June 19, 1990.
A.15. Proposed Memo to Thomas M. Novak from R. Wayne Houston on the Subject Staff Review of Proof and Review Copy of Proposed Technical Specifica-tions for McGuire Units 1&2.
06/06/1983.
A.16. Memorandum for Darrel G. Eisenhut, from:
Robert M. Bernero, " Concerns on McGUIRE Technical Specifications," dated August 30, 1984.
3-49 i
Distribution://
Docket file NkC/POR LPDR T.M. Murley/Miraglia J.M. Taylor J.H. Sniezek D.B. Matthews D.S. Hood A.C. Thadani K.D. Desai H.I. Smith (4 Copies)
R.B.A. Licciardo (2 Copies)
Chairman K.M. Carr Comissioner K.C. Regers Comissioner F.J. Remick Commissioner J.R. Curtiss t
9 4
Mvh
t ENCLOSURE 2.
,,,, u,, 'e, i
UNIT ED $TATts f,
f Jg..r..
,- t NUOLE AR REGULATORY COMMlf.SION viA s e ct o. c. c.ressa j
s, y
Jun,., b. 4
- ...a MEM%Ah0Vw. FOR:
Erian k'.
S h e *: r., CMef Rea:::* Systers E*an:t
. Division cf Systems Inte; ration FROM:
Robert E. A. Licciarco hu: lear Engineer Ret: tor Systers Branch Division of Systems 2nte; ration
$UEJE*T:
REVIEW' 0F.M GU RE TECHN20AL $PECIFICAT}ONS
REFERENCE:
a) Meme from Harcic R. Denten, Directer Office ef hu: lear Reacter Reguiad en for Darrell G. Eisenhut, Dire:ter Division cf Licensing an, Ro;te J. Mattson, Directer Oivision of Systers Inte;*atien on the
Subject:
DIFFERINS FROFE5520NAL 0F]NION OF MR. LIC !ARDO REGAR;ING M*GUIRE TECHN10AL 5FEC2F2tATION and catec: March 21, 1954 b) Meme f rom Erian k', Sheron, Chief, R$E, O!! to Robert Licciarcc R$!, 051 cate: April li, 195 cn the
Subject:
M;GUIRE TECHN} CAL SFECIFICAi!ONS A552 GNv.ENT
\\
i referen:e your meme te referen:e D) recuesting review of the M:Cuire Technical Specifications to an ac:eptatie fermat, in response te the reevirement ef ref erence a) f or a coordinatec review cf the cen: erns arising free the writer's ear'iier OPD.
Please finc attached cepy Cf a CC:vient endtle: "McGuire Units 1 l. 2:
Preposed Technical Specifications; Review cf Proef anc Revie. *c;y,6 which is in response to ycur request.
The revie. is com?csec cf two sections.
The first se: tion is entitled " Pre Revie. } nf ormation" which ettails the Easis, Furpose an* Resources, Schetele, Evaivatien Meth0c, Reguistory Reevirements and Licensing Cchsecuences of the Review.
The seconc se:tien contains the Detaile: Review.
Sin:e tne staff reovirec this cetailee review to be centuctr> without any formal, er substantive informal ciscussion, both within en: withcut, R$5, 2 presume that it is to be usec as a casis f er tne tcercinatien statet in Harcic R. Dent 0n's letter to ref erente 4), namely that "The Divisien Cf Systems Integration, in ccercination with OL, shall have pecpie that are knc lec;eable at:vt the technical surje:ts raiset ty Mr. Li::iar::, the standere technical spe:ificatiens, an: the M:Guire technical spe:ificatiens
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- riter consice*s that su:t & ce:*cinate: rev4e. in: 1v:in; c:rst u:th e :ridese is an essentiti consecuen:e cf any sv:h co:vnent The writer 6150 believes that such ccnstruttien Essi be evelese: en tne basis cf resp *.sitie written an: signe: comment within int Regulatory Framewerk.
The writer wCv1C be please: te participate in this ec:rcination as repvire:.
The write
- is autre that RSE staff has received copies of ths writer's initial preposec rneme t; T. M. Novak from R. V. Hcusten en the subje:t ef:
57 AF F REVIEW OF PR;;r AN; Rty3E * :opy 0F PROPO$ED TE;HNICAL SFIC;F10ATIONS FOR M GUIRE UNITS 1 & 2" cate: 06/15/E2, an: threv;b this actien is please: to have mace an ently contributien te re ent reviews cf Techni:41 Spe:ifications for Operating License A;plications.
Further, the writer has been infortned that the above reference: them: (cf Of/15/63) *ts also provice: to Westinghouse (h') and n:tes t.: sut$ecuent Ctveloptt.ents cf significan:e:
1)
In respenst to a cuestien f rem M. Wig:0r cen:erning "Vertie," en "Oc1:
Overpressure Mitigation', W het new recen*1y sved ttee e ie:i:t1 re;crt entitle: " Celt Overpressure Mitigating Systems,' cate: Fttruary 1964, for tevie by NRO.
2)
W has recen*1y revie.e: its position en Reteter Coolant System (RC$)
3perability reevirements in MODE 3 an: f rca this hts cete*:rine: the neef fer ac:itienti operatie R $ pumps ever those requiret in the W Si! fer the case cf "Uncontrellec Reg Civster Centrcl Assettly Etnk WIth:ra.ti F rom a $vb:ritical Condition."
E,cth cf the above items 1) and 2) wert the tsutje:t cf spe:ific concern in the reference: rnem prepose: Dy the writer, anc it is en:curaging te note the early i
response by t' tc inese safety issues.
Y.w.c w
R. B. A. Licciarce O!STRIBUTION
Attachment:
As stated Central Fue REB R/F cc:
H.R. Denten RLiccited: R/F R. Mattson gt jee 4,7e3 Op; p4),
R. V. Hovsten w/tttachmen RLi::iarco N.
Leuden w/ettechment l
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