ML20236A481
| ML20236A481 | |
| Person / Time | |
|---|---|
| Site: | Yankee Rowe |
| Issue date: | 10/31/1987 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| References | |
| NUREG-0825, NUREG-0825-S01, NUREG-825, NUREG-825-S1, NUDOCS 8710220292 | |
| Download: ML20236A481 (34) | |
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NUREG-0825 Supplement No.1-- -l LIntegrated Plant Safety Assessment Systematic Evaluation Program Yankee Nuclear Power Station Yankee Atomic Electric Company Docket No. 50-29 a.
U.S. Nuclear Regulatory Commission
. Office of. Nuclear Reactor Regulation October 1987
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NOTICE Availability of Reference Materials Cited in NRC Publications
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NUREG-0825 Supplement No.1- .
l Integrated Plant Safety Assessment .
Systematic Evaluation Program i
l Yankee Nuclear Power Station )
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- Yankee Atomic Electric Company j Docket No. 50-29 {
U.S. Nuclear Regulatory Commission Office of Nuclear Reactor Regulation October 1987 f e "%g
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,- 1 ABSTRACT
.The.V.S. Nuclear; Regulatory Commission'(NRC) has prepared Supplement l' to the .
-final-Integrated Plant' Safety Assessment Report (IPSAR)-(NUREG-0825), under the
, scope of the Systematic. Evaluation Program (SEP), for~ Yankee Atomic Electric 4 r : Company's-Yankee Nuclear Power Station located in Rowe, Massachusetts. -The SEP was; initiated by the~NRC to review the design of, older operating. nuclear-power-plants to reconfirm andi document their. safety. This report documents the review completed-under the SEP:for those issues that required refined engineering evalu-ationscor the continuation of ongoing evaluations after the Final IPSAR for the
' Yankee plant was: issued. The review has provided for (1)~an assessment of the significance of: differences between current technical positions on selected safety issues and those that existed.when Yankee was licensed, (2) a basis for- '
deciding how these. differences should be resolved in an integrated plant review,
'and (3) a documented eviluation of plant' safety, i I
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Yankee SEP iii
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TABLE OF CONTENTS Page ABSTRACT............................................................... iii ACRONYMS AND INITIALISMS............................................... ix 1 INTRODUCTION...................................................... 1-1 2 TOPICS THAT REQUIRED REFINED ENGINEERING ANALYSIS OR CONTINUATION OF ONG0ING EVALUATION................................ 2-1 2.1 Topics II-3.B, Flooding Potential and Protection Require-ments; II-3.B.1, Capability of Operating Plants To Cope With i
Design-Basis Flooding Conditions; and III-3.A, Effects of 1
. High Water Level on Structures............................... 2-1 2.2 Topic III-1, Classification of Structures, !
Components, and Systems (Seismic and Quality)................ 2-1 2.3 Topics III-2, Wind and Tornado Loadings, and III-4.A, Tornado Missiles.................................... 2-2 2.4 Topic III-5.A, Effects of Pipe Break on Structures, Components, and Systems Inside Containment................... 2-3 2.5 Topic III-6, Seismic Design Considerations................... 2-4 2.6 Topic III-7.8, Design Codes, Design Criteria, Load Combinations, and Reactor Cavity Design Criteria............. 2-4 2.7 Topic VI-1, Organic Materials and Postaccident Chemistry.................................................... 2-5 1 2.7.1 Sump Water Chemistry.................................. 2-5 2.8 Topic VI-4, Containment Isolation System. . . . . . . . . . . . . . . . . . . . . 2-5 2.9 Topic VIII-3.B, DC Power System Bus Voltage Monitoring and Annunciation.................................. 2-5 ;
2.9.1 Battery Current / Discharge and Fuse Open Alarm......... 2-5 I
2.10 Topic VIII-4, Electrical Penetrations of j Reactor Containment.......................................... 2-6 l
2.10.1 Low-Voltage Penetrations............................. 2-6 3 IPSAR TOPICS RESOLVED BY CHANGES TO PLANT TECHNICAL SPECIFICATIONS.................................................... 3-1 l 3.1 Topic VI-7.A.3, Emergency Core Cooling System Actuation System....................................................... 3-1 3.2 Topic VI-10.A Testing of Reactor Trip System and Engineered Safety Features, Including Response-Time Testing............. 3-1 3.3 Topic XV-19. Loss-of-Coolant Accidents Resulting From Spectrum of Postulated Piping Breaks Within the Reactor Cool a nt P re s su re Bou nda ry. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3-1 Yankee SEP v
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TABLE OF CONTENTS (Continued)
?; , .Page 4 -IPSAR TOPIC RESOLUTIONS CONFIRMED BY NRC REGION I 0FFICE............................................................. 4-1
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E4.1' Topics II-3.B, Flooding Potential and Protection Require-ments; II-3.B.1, Capability of Operating. Plants To Cope With' Design-Basis Flooding. Conditions; and II-3.C,
- Safety-Related Water Supply (Ultimate Heat Sink (VHS)). . . . . . . . 4-1 4.1.1 ' Roof Flooding.......................................... 4-1 l 4.2 Topic III-3.C,nInservice Inspection of Water - ,
Control Structures............................................ 4-1 i
.4.2.1 Inspection Program for Harriman and Sherman Dams................................................... 4 l 4.2.2 Inspection Program for YAEC Water-
' Control. Structures..................................... 4-l' )
1 4.3 Topic III-5.B, Pipe Break Outside Containment................. 4-1 !
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4.3.1 Main Steam Line Break.................................. 4-1 j 4.3.2 Jet Impingement on Switchgear Room Wall................ 4-2 '
E 4.4 Topic III-10.A, Thermal-Overload Protection for Motors of Motor-0perated Va1ves............................... 4-2 4.4.1 Bypass of Thermal-0verload 0evices..................... 4-2 4.5 Topics V-10.B, Residual Heat Removal System Reliability; V-II.B,~ Residual Heat Removal System i Interlock Requirements (Systems); and VII-3, 1 Systems Required for Safe Shutdown (Systems).................. 4-2 l 4.5.1 Shutdown Cooling System'0verpressurization............. 4-2
.4.6. Topic VI-1, Organic Materials and Postaccident Chemistry..................................................... 4-2 4.6.1 Surface Coatings Inspection Program.................... 4-2
.4. 7 Topic VI-4, Containment Isolation System...................... 4-2 4.7.1 Low-Pressure Surge Tank (LPST)......................... 4-2 4.8 Topic VIII-1.A, Potential Equipment Failures Associated With Degraded Grid Voltage......................... 4-3 Yankee SEP vi l
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TABLE OF CONTENTS (Continued)
Page 4.9 Topic IX-5, Ventilation Systems.............................. 4-3 4.9.1 Diesel Generator Building Ventilation System............................................... 4-3 l APPENDIX--NRC STAFF CONTRIBUTORS AND CONSULTANTS
?Y Yankee SEP vii
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t ACRONYMS'AND INITIALISMS
'CFR. - Code of Federal Regulations LG00 . general design criterion (a)
IEEE= Institutehof Electricaliand Electronics Engineers Integrated Plant Safety Assessment Report LIPSAR~
- ISI . . inservice:-inspection LOCA loss of-coolant accidint LPST ' low pressure surge tank NRC ' U.S. Nuclear Regulatory Commission' PRA- - probabilistic risk assessment RG . : regulatory guide . ..
-SEP. Systematic Evaluation Program- d.
SER - safety evaluation report j
,SRP- Standard Review. Plan- l TS -Technical Specification (s) 'I UHS ultimate ~ heat sink i
- YAEC Yankee Atomic Electric Company.
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Yankee SEP ix 1
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.r INTEGRATED PLANT SAFETY. ASSESSMENT REPORT SUPPLEMENT N0. 1 SYSTEMATIC EVALUATION PROGRAM YANKEE NUCLEAR POWER STATION:
4 1 INTRODUCTION
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The Systematic Evaluation-Program (SEP) was. initiated by the U.S. Nuclear Regulatory Commission (NRC) to review the designs of older operating nuclear
- power plants to reconfirm and document'their safety. The review provides (1) an assessment'of the significance of differences between current technical ;
positions on safety issues and those that existed when a particular plant was '
- licensed, (2)-a basis for deciding how these differences should be resolved in an integrated _ plant review, and (3) a documented evaluation of plant safety.
- The_results of.the SEP review of Yankee were published in NUREG-0825, the Final
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Integrated Plant Safety Assessment Report (IPSAR), dated June 1983. The review compared the as-built plant design with current review criteria in 137 dif-ferent' areas defined as " topics." During the review, 48 topics were deleted from consideration in the SEP because a review was being conducted under other
- programs (unresolved safety issue or Three Mile Island Action Plan' tasks), the topic was not applicable to' Yankee, or the items to be reviewed under that topic did not exist at the site.
Of the original 137. topics, 89 were, therefore, reviewed for Yankee; of these, 51 met current criteria or were acceptable on another defined basis. From the review of the 38 remaining: topics, certain aspects of plant design were found to differ from current criteria. _The integrated assessment consisted of evaluating the safety significance and other factors of the identified dif-ferences from current design to arrive at decisions about whether modification i war necessary from an overall plant safety viewpoint. To arrive at these de;isions, engineering judgment was used as well as the results of a limited prebabilistic risk assessment study. l i
In general, the staff's positions in the integrated assessment fell into one '
or more of the following categories: (1) equipment modification or addition, (2)' procedure development or Technical Specification changes, (3) refined engineering analysis or continuation of ongoing evaluation, and (4) no modifi-cation necessary. Table 4.1 of the IPSAR summarizes the staff's integrated i assessment positions and documents the licensee's agreement with those i positions. I For those positions classified as either Category (1) or (2), the IPSAR lists the scheduled completion dates agreed upon by the staff and the licensee. !
Region I has verified the implementation of these positions or identified cer- I tain corrective actions required, as described in Section 4 of this report. )
l Yankee SEP 1-1 L !
For those positions classified as Category (3), the licensee has provided the results of its evaluation or engineering analyses. The purpose of this supple-ment to the IPSAR is to provide the staff's evaluation of the Category (3) is-sues and to summarize the status of all actions to be implemented as a result of the SEP review. In those cases where analyses are continuing, the staff's evaluation identifies the analyses to be performed and the acceptance criteria that will be used to design the optimum plant modifications, if necessary. Any pre-implementation staff reviews required for these ongoing analyses, after this supplement has been issued, will be summarized in individual safety evalu-ation reports as the analyses are completed.
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2 TOPICS THAT REQUIRED REFINED ENGINEERING ANALYSIS OR CONTINUATION OF ONGOING EVALUATION Table 2.1 of this report presents a list of all issues that were evaluated in ;
the IPSAR. The licensee has submitted an evaluation of each of the items identified in the final IPSAR as requiring additional analysis. A summary of the staff's findings of these items is presented in Sections 2.1 through 2.10 below. Each section references the staff's Safety Evaluation Report, if appli-cable, which provides more detail regarding the basis for the staff's conclu-sions. Factors considered in reaching a staff conclusion for each item include the perceived safety. significance of the difference from current licensing criteria and a qualitative assessment of the financial and radiation exposure costs to make a modification. The evaluation of these issues also considered ,
, any applicable risk perspectives, developed for the integrated assessment and described in the IPSAR, and related corrective actions proposed by the licensee I as part of the integrated assessment or as a result of the subsequent evaluations.
~ 2.1 Topics II-3.B, Flooding Potential and Protection Requirements; II-3.B.1, Capability of Operating Plants To Cope With Design-Basis Flooding Condi-tions (NUREG-0825, Section 4.1); and III-3.A, Effects of High Water Level on Structures (NUREG-0825, Section 4.6)
General Design Criterion (GDC) 2 in Title 10, Part 50 of the Code of Federal Regulations (10 CFR 50), as implemented by Standard Review Plan (SRP) Sec-Fion 2.4.5 (NUREG-0800) and Regulatory Guide (RG) 1.59, requires that struc-tures, systems, and components important to safety be designed to withstand the effects of natural phenomena such as floods.
A failure of the Harriman Dam from the effects of probable maximum precipitation constitutes the only flooding threat to the site. The Federal Energy Regula-tory Commission is responsible for the review of the integrity of this dam.
The staff will receive the results of this review by the end of 1987 and will use that document to close out the NRC review of this concern.
2.2 Topic III-1, Classification of Structures, Components, and Systems (Seismic and Quality) (NUREG-0825, Section 4 4) 10 CFR 50 (G0C 1), as implemented by RG 1.26, requires that structures, systems, and components important to safety be designed, fabricated, erected, and tested to quality standards commensurate with the importance of the safety functions to be performed.
In IPSAR Section 4.4, the staff concluded that the licensee should supply addi-tional information to address the following areas:
(1) radiography (2) fracture toughness (3) valves (4) pumps Yankee SEP 2-1
(5) storage tanks (6) piping (7)- codes and standards (8) pressure vessels The licensee proposed to evaluate the safety significance of the components.
and systems in question and show that they are adequately monitored by a formal inspection program or that the risk from failure is negligible. The results of the licensee's review were provided in a letter dated September 26, 1984.
The staff, in its evaluation issued on August 25, 1986, concluded that these issues were satisfactorily resolved.
2.3 Topics III-2, Wind and Tornado Loadings (NUREG-0825, Section 4.5), and III-4.A, Tornado Missiles (NUREG-0825, Section 4.8) 10 CFR 50 (GDC 2), as implemented by SRP Sections 3.3.1 and 3.3.2 and RGs 1.76 and 1.117, requires that the plant be designed to withstand the ef fects of natural phenomena such as wind and tornados.
In IPSAR Sections 4.5 and 4.8, the staff stated that some structures and compo-nents important to safety would not withstand the 10-7/ year tornado that was recommended in the evaluation for SEP Topic II-2.A. l l
4 In the IPSAR, the licensee proposed a 10-5/ year tornado (median) of 110 mph as !
a more appropriate design basis and proposed to satisfy the following objec- i tives to demonstrate adequate tornado protection, rather than to upgrade the l plant to protect against the 10-7 tornado: 1 (1) Maintain integrity of the reactor coolant pressure boundary.
-(2) Maintain integrity of the secondary system pressure boundary as a heat sink.
(3) Ensure capability for steam generator feedwater and primary system makeup.
The staff concluded that the general method was acceptable but that the follow-ing specific recommendations should be followed:
(1) Determine the capacity of systems, structures, and components raquired for reaching a hot shutdown condition at the 10-4/ year and 10-5/ year upper 95%
confidence limit windspeeds.
(2) Determine the modifications needed to upgrade the plant to protect against both windspeeds.
(3) Estimate the costs of such modifications.
(4) Perform a cost-benefit evaluation to decide what modifications to make.
For tornado missiles, the staff's position in the IPSAR was that a steel rod and a utility pole should be considered in the licensee's analysis of the effects of winds and tornados.
The licensee submitted the cost-benefit evaluation for wind and tornados that stated that certain plant modifications should be implemented to improve plant capability to withstand such events. These include modifications to l
Yankee SEP 2-2 E_
(1) the block walls in the turbine buil' ding and primary auxiliary building (2) the cable spreading room (3) the main steam /feedwater piping and support structure (4) the diesel generator building west wall The staff, in its evaluation issued on May 13, 1987, concluded that with the implementation of the specified modifications, the risk from high wind / tornado events'(including tornado generated missiles) is acceptably low.
2.4 Topic III-5.A, Effects of Pi)e Break on Structures, Components. and Systems Inside Containment (4UREG-0825, Section 4.9) 10 CFR 50 (GDC 4), as implemented by SRP Section 3.6.2, requires, in part, that structures, systems, and components important to safety be appropriately pro-tected against dynamic effects such as pipe whip and discharging fluids.
In IPSAR Section 4.9, the staff identified four areas requiring further evalua-tion. These areas were related to (1) clarification of assumptions used in the jet impingement and pipe whip evaluations (2) thrust forces on steam generator (3) blister 12E (electrical penetration) l (4) loop compartment walls Each of these issues was resolved as discussed in the following sections.
Jet Impingement and Pipe Whip Evaluations In a letter dated September 27, 1934, the licensee stated that two areas would 1 require further assessmerit on the basis of its reevaluation of the effects of jet impingement.
The first area was the effect of a break in the 5-inch crossover piping. Such a break could have unacceptable consequences. Therefore, the licensee performed a leak-before-break analysis to demonstrate that a rupture would not occur.
This analysis was provided in a letter dated October 1, 1986. The staff, in its evaluation issued on July 16, 1987, concluded that the licensee's analyses were acceptable contingent on the licensee's commitments to (1) modify the pro-cedure for visual inspection of 5-inch crossover piping for potential inservice pipe degradation, (2) perform augmented inspection of eight main steam piping welds, and (3) modify steam generator blowdown piping supports for jet impinge- l ment loads. '
The second area was a break of steam generator blowdown piping where the four lines enter the containment. The licensee has committed to include jet impinge-ment loads in the seismic reanalysis of this piping, so that the piping line would not fail as a result of the ef fects of a break in an adjacent line. The l staff finds this commitment acceptable.
! Yankee SEP 2-3 i
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Thrust Forces on Steam Generator The. licensee analyzed the effects'of a rupture of a main steam line at the noz-zie on the steam generator in a submittal dated March 26, 1984. The staff found
.that the structural integrity of the generator was acceptable. In a letter dated March 16, 1986, the staff raised a question concerning a horizontal break at the main steam outlet. In its October 1, 1986 submittal, the licensee stated i that the steam generator could not withstand the loading and proposed an aug- i mented inservice inspection (ISI) program for these welds. By letter dated l July 16, 1987, the staff concluded that this resolution was acceptable.
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. 1 Blister 12E (Electrical Penetration)
! To resolve the concern regarding the effects of jet impingement on ele ~ctrical penetration blister 12E, the licensee proposed an augmented ISI program for the 1 welds. By letter dated May 3, 1984, the staff concluded that this resolution 3 t
was acceptable, Loop Compartment Walls The effects of pipe breaks in large reactor coolant system piping are' covered !
by the analyses in WCAP-9558, which provides the technical basis for not ;
postulating double-ended pipe breaks in this piping. As discussed in the let-ter dated May 3, 1984, the staff finds this analysis acceptable to resolve this l
-concern. j l
L 2.5 Topic III-6, Seismic Design Considerations (NUREG-0825, Section 4.11) 10 CFR 50 (GDC 2) and 10 CFR 100, Appendix A, as implemented by SRP Sections 2.5, 3.7, 3.8, 3.9, and 3.10 and SEP' review criteria (NUREG/CR-0098), require that structures, systems, and components important to safety be designed to withstand the effects of natural phenomena such as earthquakes.
In the IPSAR, the staff stated that the licensee's analyses of piping and major mechanical equipment were not complete. There also were some open issues per-taining to structures.
By letter dated July 16, 1987, the staff issued its safety evaluation of the seismic reevaluation program for Yankee. The licensee has made specific commit-ments for plant modifications and further analyses as a result of this review.
The staff finds these commitments acceptable.
- 2. 6 Topic III-7.8, Design Codes, Design Criteria, Load Combinations,ind Reactor Cavity Design Criteria (NUREG-0825, Section 4.12) l 10 CFR 50 (GOC 1, 2, and 4), as implemented by SRP Section 3.8, requires that l structures, systems, and components be designed for the loadings they may experi-l ence and that they conform to applicable codes and standards, in a letter dated July 16, 1987, the staff requested that the licensee reanalyze the column to vapor container shell connections; therefore, this remains an open item. The licensee agreed to redeck the heating boiler room roof and the lower roof of the primary auxiliary building to resolve a concern about snow loads.
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As discussed in IPSAR Section 4.12, the licensee pro' posed to perform, on a sam- !
pling basis, an evaluation of the code, load, and load combination issues de-lineated by the staff in order to assess the adequacy of as-built structures at Yankee. By letter dated December 4, 1986, the licensee supplied the results of its review. In addition, the staff requested information regarding the effects of snow loading on plant structures. The licensee responded in a letter dated September 8, 1986. The staff's review of both issues is documented in a letter dated July 16, 1987, which'found the licensee's analyses acceptable.
2.7 Topic VI-1, Organic Materials and Postaccident Chemistry (NUREG-0825, Section 4.21) 2.7.1 Sump Water' Chemistry
'10 CFR 50 (GDC 14) requires that the reactor coolant pressure boundary be de-signed so that it has a low probability of abnormal degradation or rapidly prop-agating failure.
A low pH value increases the potential for stress-corrosion cracking of piping systems. 'Following an accident, the pH of the sump water would be low because of the presence of boric acid and hydrochloric acid (formed by radiolysis).
In IPSAR Section 4.21.1, the licensee agreed to provide a means for control-ling the pH of sump water following a loss-of-coolant accident. The licensee installed trisodium phosphate baskets in the sump during the 1985 refueling out-age. As discussed in a staff evaluation dated November 19, 1984, the staff finds this resolution acceptable.
2.8 Topic VI-4, Containment Isolation System (NUREG-0825, Section 4.22) 10 CFR 50 (GDC 54, 55, 56, and 57), as implemented by SRP Section 6.2.4 and RGs'1.11 and 1.141, requires isolation provisions for the lines penetrating primary containment to maintain an essentially leaktight barrier against the uncontrolled release of radioactivity to the environment.
In the IPSAR, the staff noted several areas that required further evaluation.
The licensee provided a response to the concerns by letter dated March 16, 1983. !
There are some penetrations that do not contain redundant isolation barriers. ;
The staff evaluated the risk reduction that would result if additional valving was installed. In an August 28, 1986 evaluation, the staff concluded that only a minimal reduction in risk would occur and, therefore, the modifications were not required.
2.9 Topic VIII-3.B, DC Power System Bus Voltage Monitoring and Annunciation (NUREG-0825, Section 4.28) 2.9.1 Battery _ Current / Discharge and Fuse Open Alarm !
In IPSAR Section 4.28.1, the staff recommended that an ammeter be installed to indicate battery current for charge / discharge. However, the licensee sub-sequently performed a test that demonstrated that installation of an ammeter l
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Yankee SEP 2-5
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would not be effecttve in detecting high resistance connections. As an alter-native, the licensee has in plh e test and surveillance procedures for plant batteries to check for loose connections and for buildup of corrosion. There-fore, by letter dated September 19, 1985, the staff concluded that this issue was resolved.
l 2.10 Topic VIII-4, Electrical Penetrations of Reactor Containment (NUREG-0825, Section 4.29) 10 CFR 50 (GDC 50), as implemented by RG 1.63 and Institute of Electrical and Electronics Engineers (IEEE) Std. 317-1972., requires that penetrations be de-signed so that the containment structure can accommodate, without exceeding the design leakage rate, the calculated pressure, temperature, and other environ-mental conditions resulting from any loss-of-coolant accident (LOCA).
2.10.1 Low-Voltage Penetrations
'In IPSAR Section 4.29.2, the staff noted that some low-voltage electrical pene-trations served components inside.the containment for which qualification for the LOCA environment had not been established. These were identified as category.
B penetrations. Current criteria would require a backup to the primary protec-tion device. .In the IPSAR, the staff concluded that modifying the penetrations to add backup protection would result in only a small improvement in risk.
Therefore, the staff concluded that modification was not necessary provided the licensee determined that the existing circuit protection met or exceeded that assumed in the staff's risk assessment.
In Appendix D to the IPSAR, the staff evaluated the risk significance of the category 8 penetrations. In the assesscent, it was assumed that a penetration fault occurs if an electrical fault (circuit overload) exists and the breaker fails to isolate the circuit. The breaker fault is assumed to be an independent failure; thus, the fault clearing time of the primary protection device must be shorter than the time to reach design temperature of the penetration.
By letter dated January 4, 1984, the licensee described the protection for the category B penetrations. Since they all had primary protection devices consis-tent with the probabilistic risk assessment assumptions, the staff concluded that this issue was resolved in an evaluation issued on September 19, 1985.
Yankee SEP 2-6
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3 IPSAR 1 PICS RESOLVED BY CHANGES T,0 PLANT TECHNICAL SPECIFICATIONS integrated assessment for Yankee, a number of issues were resolved bn Duringnini by,en tht)nts fech thblicenset \ to perform evaluations in order to determine
!M whet r modificatf p[f'q'blant T4chnial Specifications were warranted.
,/.
s' u !
i si 'j?his section describes t'fa actions taken,regarding; resolution of IPSAR issues a, involving Technyal .;
Wej;fication changes. ,
t '3.1 To)ic VI-7.n.7,'Emugency Core Cooling System Actuation System
/ i (N JREG-0825,'Sn. tion 4. 23)
In Ip5AR Section 4.2T, the staff noted that the Yankee Technical Specifications (TS) allow the exclusion of testfag automatic valves in the flow path of the emcty pcy core cooling system. Therafore, the staff recommended that the phrase "Exc1 ding Autom9 tic" be deleted frc= tne TS. This change was submitted by the licensee by letter dated January 23,0.984 and apprcved in Amendment 83 to the license on July 1,1985.. 4
+
t
- 3. 2 Topic VI-l'O. A, Testing of Reactor Trir. St; tem ar.d Engineered Safety Features,',IncludinojesponseTimeTestinGNUREG-0825,Section4.24)
InIPSAR3sction4.24,th4staffdiscussedresponse-timetestingofreactor s
protection Nid engirjeered shfeguard fqa,tures. The st.aff recommended that test-
~
ing of respon'se ' Ants of important camponents now addressed by plant procedures be included in the)% By letter dated January' 23,1984, a TS change request was submitted by,the licensee. In Amendment G& to the license, issued on July 1, 1985, response-t19e testbg of dieael generator starting was approved.
J .' 3 Topic XV-19) Lovd-o$-CoolanMccidents Resulting From Spectrum of Postu' c.te dijiv.g Breaki~IquM the Reactor Coolant Pressure Boundary _
-,- fKURUD2K 5ecti,onn 4.3ST y ;
l ,j '
3 +
In IPSAR'Section 4.35, the staff ider,tified a concern that calculated doses following a LOCA might exceed 10 CFR 100 gaidelin.et. The postulated 1 gpm leakage of recirculated core c~ooling water outside the containment was a major ifactor. Therefore, the staff concluded ti.ut the licensee should limit the l' leakage by TS so that doses following a LOCA bould satisfy the guidelines. By letter dated May 7,1985, the licensee submitted a proposed TS change to include
\ a leakage limit of 50 gallons per day for the recirculation system.
' The proposed TS limits were approved in Amendment 90 to the license, dated December 16, 1985.
/ ! ;
)' /
i 1
YankeeSEh / 3-1
-__--lN_____-_-______________ 1__-!____-_'_'--___----_------_--_--------------------------------
i n:
~
-4' IPSAR TOPIC RESOLUTIONS CONFIRMED BY NRC REGION I 0FFICE During;the integrated assessment for' Yankee, a number of issues were resolved zby: commitments made by the licensee for specific plant modifications or pro-
- cedural changes. After'the IPSAR for Yankee was issued, the Region I office was asked through Task Interface Agreement 83 to verify that plant modifications y had been implemented and to review changes to plant operating procedures made '
by the l_icensee. Table 4.1 provides a list of IPSAR actions for which confir-mation by the Region I office.was requested.
Region I: personnel conducted onsite inspections for each item' identified in !
Table 4.'1.~ The inspections consisted-of examinations of installed equipment as well as a review of supporting procedures and'other documentation. The Region I i office concluded that the licensee had met the commitments documented in the '
IPSAR'for the items in Table 4.1. Inspection. findings with the results of the review are documented in inspection reports as noted in the following sections.
4,1 . Topics II-3.8, Flooding Potential and Protection Requirements; II-3.B.1, Capability of Operating Plants To Cope With Design-Basis. Flooding Condi-tions; and Il-3.C, Safety-Related Water Supply (Ultimate-Heat Sink (UHS))
(NOREG-0825, Section 4.1).
- )
4.1.1 . Roof Flooding I L
As. discussed in IPSAR Section 4.1.5, the licensee installed scuppers on the turbine building roof. (Inspection Report 83-15) 4.2 ' Topic III-3.C, Inservice Inspection of Water-Control Structures (NUREG-0825c Section 4.7) 4.2.1 Inspection Program for'Harriman and Sherman Dams
,As discussed in IPSAR Section 4.7.1, the licensee is retaining copies of l Federal Energy Regulatory Commission inspection reports on the Sherman and l' Harriman Dams. (Inspection Report 83-15) 4.2.2 Inspection Program for YAEC Water-control Structures '
-As discussed in IPSAR Section 4.7.2, the licensee has implemented a formal in- !
spection program for water-control structures. (Inspection Report 84-20) 4.3 Topic III-5.B, Pipe Break Outside Containment (NUREG-0825, Section 4.10) 1 4.3.1 Main Steam Line Break !
~As discussed in IPSAR Section 4.10.1, the licensee has modified the inservice inspection program to include augmented inspection of welds on the steam lines.
These welds were examined by ultrasonic and magnetic particle techniques. (In- !
spection Report 84-01)
'l Yankee SEP 4-1 i'
4.3.2 Jet Impingement on Switchgear Room Wall As discussed in IPSAR Section 4.10.2, the licensee committed to install a shield plate on the switchgear room wall to protect equipment from the adverse ,
effects of a pipe break. Installation is complete. (Inspection Report 85-04) )
4.4 Topic III-10.A, Thermal-Overload Protection for Motors of Motor-Operated Valves (NUREG-0825, Section 4.14) 4.4.1 Bypass of Thermal-Overload Devices As discussed in IPSAR Section 4.14.1, the licensee performed a one-time test of the thermal-overload setpoints for motor-operated valves to determine whether the setpoints were adequate.
As discussed in Inspection Report 84-20, the valves that were tested were determined to be acceptable; however, eight other valves listed in Technical Specification Section 4.5.2 were not tested. By letter dated April 24, 1985, the licensee noted that these valves were part of other plant modifications and the overloads were functionally tested as part of their installation. During the following refueling outage, the thermal overload setpoints for these eight valves were tested to close out this issue. (Inspection Report 85-11) !
4.5 Topics V-10.B, Residual Heat Removal System Reliability; V-II.B, Residual !
Heat Removal System Interlock Requirements (Systems); and VII-3, Systems Required for Safe Shutdown (Systems) (NUREG-0825, Section 4.19) i 4.5.1 Shutdown Cooling System Overpressurization As discussed in IPSAR Section 4.19.5, the licensee has installed an interlock on one valve on both the inlet and outlet of the shutdown cooling system. The licensee also has completsd operations and surveillance test changes reflecting t this modification. (Inspection Report 84-20) 4.6 Topic VI-1, Organic Materials and Postaccident Chemistry (NUREG-0825, Section 4.21) 4.6.1 Surface Coatings Inspection Program As discussed in IPSAR Section 4.21.2, in a letter dated April 24, 1985, the licensee described its inspection program for containment coatings that has (Inspection Report 85-11) j been implemented at the plant. l 4.7 Topic VI-4, Containment Isolation System (NUREG-0825, Section 4.22) 4.7.1 Low-Pressure Surge Tank (LPST)
As discussed in IPSAR Section 4.22.6, the licensee completed the modifications to remove the low pressure surge tank as an extension of the containment in j j
1984. (Inspection Report 84-20) l Yankee SEP 4-2
i I
4.8 Topic VIII-1.A, Potential Equipment Failures Associated With Degraded Grid )
Voltage (NUREG-0825, Section 4.27) i 1
As part of the' resolution of the degraded grid voltage multiplant issue, the l licensee committed to develop procedures for diesel generator load shedding. '
In a. letter dated March 11, 1985, the staff concluded that the procedures provide the necessary protection of the Class 1E electrical system from a degraded grid voltage condition (when there is no loss-of-coolant accident).
4.9 Topic IX-5, Ventilation Systems (NUREG-0825, Section 4.31) ;
4.9.1 Diesel Generator Building Ventilation System As discussed in IPSAR Section 4.31.3, the licensee modified the ventilation system for the diesel generator building to eliminate single-failure vulner-abilities. The modifications were completed during the 1984 refueling outage.
(Inspection Report 84-20)
Table 4.1 Items for confirmation by NRC Region I office Item IPSAR no. Description section (1) Install improved roof drainage. 4.1.5 (2) Retain copies of Federal Energy Regulatory Commission inspection reports. 4.7.1 (3) Develop and implement formal inspection program ,
for water-control structures. 4.7.2 (4) Include welds of main steam line at non return valves in inservice inspection program. 4.10.1 (5) Install jet impingement shield plate. 4.10.2 (6) Perform one-time test of thermal-overload setpoints. 4.14.1 (7) Install pressure interlock on shutdown cooling system valves. 4.19.5 (8) Develop and implement program for inspection of containment coatings. 4.21.2 (9) Make low pressure surge tank modifications. 4.22.6 (10) Develop procedures for diesel gener-ator load shedding (degraded voltage). 4.27 (11) Modify diesel generator building ventilation system. 4.31.3 Yankee SEP 4-3
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~ APPENDIX
- 1 NRC-STAFF. CONTRIBUTORS AND CONSULTANTS.< {
I This supplement is'.a' product:offthe NRC staff and its consultants.' The NRC staff members listed below were. principal: contributors to this' report. ,A list-of.. consultants'follows the, list of staff members.
NRC STAFF j
.i -l M.LBoyle P. Chen '
-T. Cheng.
M.'. Fields .!
C. Grimes- _
.E..McKenna :
- 1
. CONSULTANTS
.I Name Affiliation. 1
~
M. Russell EG&G-Idah'o (INEL)
.A.f0kaily Franklin Research Center S. Triolo Franklin Research Center
< L, Shieh : Lawrence Livermore National Laboratory .)
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, Yankee SEP 1 Appendix l
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Supplement No. 1 J Leave o<a* (
) Tif t,5 ANO SU6 TITLE 4 H4Cs*>tNf5 ACCESS NUM8ER Integrated ant Safety Assessment f
, 04, , ,,, ops,,u ,, o Systematic E luation Program ocN ,,,
Yankee Nuclea Power Station l,,A, Se ber l987
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'p.oNin lveAn l October 1987 i
. 9 PHOJECitTA$KmQHK VNIT NUM8tR 8 P(RFORMING OHGANi2 A riON NAME AND sung AOoets$ flutuJr le Co*> -
Division of Reactor rojects I/II ..-
i Office of Nuclear Rea tor Regulation io eiN Nuuna U. S. Nuclear Regulato ' Commission .
Washington', D. C. 20555 11 $PONSURING OMG ANilAliON NAMi ANO MAILING AD- E SS fractum /# Codel I
1 24 T VPE OF REPOR T Same as 8, above Technical Report
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June 1983 - September 30, 1987
,3 sue,.uunrAnv Noris g:
Docket No. 50-29 L to ASS TM AC T (20D *ords or 88ul '
1
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1 1The U. S. Nuclear Regulatory Commissi . (NRC)has re>ared Sup Integrated Plant Safety Assessment Rapo t (IPSAR) p(NUREG-0825)plement , under the scope 1 to the final j of the Systematic Evaluation Program (SEP , ,
Nuclear Power Station located in Rhe )Ma or Yankee Atomic Electric Compdny's Yankee i achusetts. The SEP was initiated by the ;
NRC to review the design of older /oper,atin nuclear power plants to reconfirm and document their safety. This rep 6tt documen the review completed under the SEP for I those issues that required refi/ied engineeri evaluations or the continuation of i ongoing evaluations after the final IPSAR for -he Yankee plant was issued. The review has provided for (1) an assessment of the sign- icance of differences between current technical positions on sele ed safety issues a licensed, (2) a basis for ciding how these dif those that existed when Yankee was integrated plant review, a,d (3) a documented eval ences should be resolved in an tion of plant safety.
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