ML20244E155
Text
.
Docket Nos.:
50-443 and 50-444 JUL 3 01986 MEMORANDUM FOR:
Victor Nerses, Project Manager PWR Project Directorate #5 Division of PWR Licensing-A FROM:
Charles E. Rossi, Assistant Director Division of PWR Licensing.A i
SUBJECT:
REVISED FINAL DRAFT SSER 5 FOR SEABROOK, UNITS 1 and 2 The,2chnical review branches of the Division of PWR Licensing-A have reviewed the revised pages to the final draft SSER 5 for Seabrook, as con-tained in Attachment I to this memorandum.
Subject to the comments, from the branch chiefs, in the attached marked up pages 1-2, 9-3, 11 2 and 11-3 (Attachment 2), we concur in the revisions of the pages.
_C U :'.;.;.33 Irf, I
Charles E. Rossi, Assistant Director Division of PWR Licensing-A Attachments:
As Stated cc:
V. Noonan (w/o Attmnts)
Distribution Docket File RBallard PDf5 4 "
JMilhoan V. Nerses VBenaroya C. Possi CBerlinger i
M. Rushbrook FRosa
- See previous concurrences
'FC
- PAEB*
- PAPS *
- PAF0*
- PARS *
- PAEl*
- DPLA: 0
....:............:............:............:............:...............(
_e:...........
RME :RBallard:ss :JMilhoan
- VBenaroya
- CBerlinger
- FRosa
,ossi i :
....:...........:............:............:............:............:............y...........
j ATE :7/30/86
- 7/30/86
- 7/30/86
- 7/31/86
- 7/30/86
- 7/fB86 0FFICIAL RECORD COPY i
A g 07C (c 3'15 ~)w,o4 Wl j
s
i SEABROOK SSERS PAGES WITH SUBSTANTIVE REVISIONS SINCE TECHNICAL. CONCURRENCE i-
.SECTION PAGE 1,7 (Outstandin9 Issues) 1-2 1
1.8 (Confinnatory Issues) 1-3 l
1.9 (L.icense Conditions) 1-4 2.3.3 2-2 1
3.7.3 3-5 3.9.6 3-6 3.10.1.3 3-7 3.10.2.2 3-10 3.10.2.3 3-13 4.4.5.3.
4-1 4-2 4.4.5.4 4-3 4-4 5.4.7.5 5-3 5-4 6.2.6 6-3 7.5.2.2 7-12 I
7.5.2.4 7-13 9.3.4.2 (II.B.3) 9-3 9.5.1.4 9-8 9-9 9.5.1.5 9-12 9.5.1.9 9-13 11.4.2 11-2 11-3 11.5.2 11-4 15.9.15 (III.D.1.1) 13 App M. (PSI Relief)
M-11 l
supplement that pre-i These items and the sections of th outstanding issuesa' aluation are given below.d in sections follow i k become closes eight items.
sent results of the staf f's evlisted below tha Outstanding issues listed below thaasterisk b conditions.
l tions followed by a doub e2 3 3 ** 13.3*) testing programs (2.5.5.3, 3.9.,
1icense 6**
(2) Emergency preparedness (...and inservice insp
~
Preservice 5.2.4,* 6.6. 3*)
ipment (3.11**)
(4) 9 9)
Environmental qualification of equ** 6. 2. 8, 9.3. 4. 2, 13. 3,* 14,*
(6)
TMI Action Plan items (4.4.5.4, n
(7.3.2.8) 4)
(8)
(10) Level measurement error f or saf e shutdown (7.4.2.1,** 7.4.2.
(11) instrumentation and control ystem (7.5.2.2*)
13.2.2.1) 22 (12) Radiation data management sTMI Action Plan Item I. A.1.1 (13 l ted SER (16) Shif t Technical Advisor, i ing outstanding issues and the re a
~
1 As of this supplement, the rema n ipment (3.10) sections are Stress / dynamic qualification of equ ystem (11.4)
(6)
(14) Solid radwaste management s (20) Fire protection (9.5.1.9) n resolved h
there are some items that have beenfirmatory Section 1.8 of the SER noted thattisf action This supplement closespplement I
1.8 d by the applicant.
The confirma-j These items and the sections of t essentially to the staff's saformation has not yet bee t f f's evaluation are given be ofollowed by an lv l
11 of the confirmatory items.that present results of th that is dis-t is discussed in the section The confirmatory issue listed below tstanding tory issue listed below thaa license condition.
d by a double asterisk becomes an ou cussed in the section followe becomes nt (2.1.2)
Underground transmission line easemege and vent s
- issue,
)
(1)
(12) Analysis of the containment purnalysis (6.2.1.2) to several
/
sly provided information related (13) Containment subcompartment a (7.2.2, 7.4.2.5, 7.6.7.6) i (14) Formal documentation of prev ouinstrumentat 1-2 J
a
~
(16) Main steam atmospheric relief valves (7.4.2.2)
(17) Pressurizer auxiliary spray (7.4.2.5)
(31) Diesel generator exhaust inspection and protectic, (9.5.8)
(37) Health physics organization (12.5.1)
(40) Inadvertent boron dilution (15.4.6)
(41) Systems outside containment containing radioactive material, TMI Action Plan Item 111.0.1.1 (15.9.15*)
(42) Stress / dynamic qualification of equipment (3.10**)
As of this supplement the remaining and additional confirmatory items are given below. The staff has decided that the confirmt'ory issues listed below that i
are discussed in sections followed by an asterisk may be resolved af ter core load.
(6) Loose parts monitoring system (4.4.5.3*)
(24) Non-safety loads powered from the Class IE ac system (8.3.1.4)
(33) Sampling capability for vacuum pumps during startup (10.4.2)
(45) Steam generator tube rupture (15.6.3*)
(46) Electrical power systems, general (8.1)
(47) Environmental qualification of equipment (3.11)
(48) Instrumentation and control for safe shutdown (7.4.2.1)
(49) Cable tray supports (3.7.3*)
(50) Turbine system maintenance program (3.5.,1.3*)
(51) Inadequate core cooling, TMI Action Plan Item II.F.2 (4.4.5.4*)
(52) Postaccident monitoring (7.5.2.4*)
(53) Inservice testing of pumps and valves (3.9.6*)
(54) Tests, operational procedures, and support systems (5.4.7.5*)
s For a number of confirmatory issues, the remaining action involves verification by the NRC staff that the applicant has implemented its commitments with regard to such items as equipment installation or modification, alarms or setpoints, and plant procedures or testing.
The following confirmatory issues are included in this verification category:
(43) PVORT equipment-specific issues and generic issues (3.10.2.3)
(44) Administrative procedures (13.5.1)
(45) Emergency preparedness (2.3.3) q Verification of Items (43), (44), and (45) will be accomplished as part of the f
ongoing inspection program for Seabrook conducted by NRC Region I inspection staff. The Region I inspection staff will ensure that these items are completed and will report on the status of their completion.
1.9 License Condition Items In Section 1.9 of the SER, the staff noted several issues for which a license condition may be desirable to ensure that staff requirements are met during plant operation if those requirements have not been met before the operating Seabrook SSER 5 1-3 i
u------_-_--__--
license is issued.
The license condition may be in the form of a condition in the body of the operating license, or a limiting condition for operation in the Technical Specifications appended to the license.
This supplement closes seven license condition items. These f(ems and the sec-tions of this supplement that present results of the staff's evaluation are given below.
t.icense condition items listed below that are discussed in sec-tions followed by an asterisk become confirmatory items.
The license condition item listed below that is disc.ussed in the section followed by a double asterisk becomes an outstanding issue.
(1) Turbine system maintenance program (3.5.1.3*)
(3) Detection of inadequate core cooling, TMI Action Plan Item II.F.2 (4.4.5.4,* 4.4.8)
(5) Natural circulation tests (5.4.7.5)
(8) Compliance with NUREG-0612 (control of heavy loads) (9.1.5)
(9) Postaccident sampling, TMI Action Plan Item II.B.3 (9.3.4.2)
(10) Secondary water chemistry monitoring and control (10.3.5)
(12) Shift Technical Advisor, TMI Action Plan Item I. A.I.1 (13.1.2.2, 13.2.2.1)
(14) Solid radwaste management system (11.4**)
As of this supplement the remaining license condition items are I
(4)
Inservice inspection program (5.2.4, 6.6.3)
(13) Emergency preparedness (l'3.3)
(16) Implementation and maintenance of the physical security plan (13.6)
(17) Training during low power testing, TMI Action Plan Item I.G.1 (14)
(21) Control room design review, TMI Action Plan Item I.D.2 (18)
(22) Radiation data management system (7.5.2.2) l (23) Fire protection (9.5.1.9)
(24) Systems outside containment containing radioactive material, THI Action Plan Item III.D.1.1 (15.9.15) k 1.10 Nuclear Waste Policy Act of 1982 Section 302(b) of the Nuclear Waste Policy Act of 1982 states that NRC shall not issue or renew a license for a nuclear power reactor unless the utility has signed a contract with the Department of Energy (DOE) for disposal ser-vices, or the Secretary of Energy affirms in writing that a utility is actively Seabrook SSER 5 1-4
a and any problems will be followed through the inspection process.
The descrip-tion of the operational meteorological measurements is adequate to close the outstanding issue.
During the two phase EPIA conducted by Region I staff in December 1985 and March 1986, the appropriateness and quality of meteorological data,available for use in emergency response were reviewed as part of the operational meteorological measurements program.
Inspection Report 50-443/85-32 identified a number of
" improvement items" related to meteorological data.
(The " improvement item" classification means that compliance with requirements is marginal and that i
improvements should be made.) In a followup Region I inspection, documer,ted in Inspection Report 50-443/86-18, it was noted that while the improvements have not yet been made, a description of anticipated applicant actions was pro-cid:d.
Thus, these items are being appropriately addressed by inspection ac-tions and will be completed by fuel load.
The completion of these items will be reported by Region I staff. Therefore, with these actions the staff con-cludes that the outstanding issue related to the use of meteorological infor-mation for emergency response is closed pending confirmation from Region I.
j 2.5 Geoloqy and Seismoloqy 2.5.5 Stability of Slopes 2.5.5.1 Slope Characteristics In the SER, the staff stated that two designs would be used in constructing stone revetments to protect portions of the perimeter of the plant that could be exposed to wave action during the peak probable maximum hurricane (PMH) surge.
After the SER was issued, the applicant revised the stone revetment designs and proposed three designs instead of the original two.
(See Fig-ure 2.2.) These revisions were due to changes in the site topography that occurred during construction.
The original design for revetment "B" had been based on a slope that extended to the original land elevat' ion.
During construction, however, the land area east of revetment "B" was graded to an elevation of about 13 ft above mean sea level (ms1). This higher ground provides significant flood protection and re-suits in a reduction in the height of the design waves that could impact on revetment "B" during a PMH. As a result, the applicant reduced the size and weight of the stone required for design of revetment "B."
In the SER, the staff stated that design "A" would be used for the south and southeast stone revetments. This has been revised: Design "A" will now be used for the south perimeter revetment only.
A new design, "C," will be used
)
for the southeast portion of the revetment as shown on Figure 2.2.
This new design was also due to changes in site topography that resulted in a reduc-tion in the calculated design wave height along revetment "C."
The size and
)
weight of the stones required for revetment "C" were reduced consistent with the reduced wave height.
In Section 2.5.5.1 of the SER, the staff stated that revetment design "A" would consist of a 6-ft layer of amor stones weighing between 1.5 to 3 tons each, overlying a 3-ft layer of stones weighing 300 to 600 lb each, overlying Seabrook SSER 5 2-2
normal industry practice have more than sufficient margin to absorb the seismic loads anticipated from moderate earthquakes, such as the Seabrook seismic design-basis loads, and (d) the cable tray systems at Seabrook should have more than adequate capacity to survive the design-basis SSE.
Documentation Each particular cable tray support system should have complete documentation to serve as a basis for qualification.
This documentation shall also become the design record of that cable tray support system.
The documentation should show the basis for qualification either by test, analysis, or combined test a.d analysis.
In the event of qualification by test, the applicability of test
- sts ano configuration of cable tray supports should be demonstrated.
The applicant should also provide a checksheet for each cable tray support quali-fied by test and/or envelope analysis.
The format of the checksheet should provide for the support identification number, system configurations, loading, bracing, connection, materials used for trays and supports, and support cate-gorization. For cable tray supports qualified by support unique analysis, the specific information pertinent to modeling, analysis, and design calculations should be documented. In a letter dated June 20, 1986, the applicant agreed I
to provide documentation and checksheets and advised the staff that the documen-tation was complete. The documentation shall be in a form suitable for staff L
audit.
b endment of FSAR The applicant should amend the FSAR to reflect changes as proposed in Attach-f sent 2 to the applicant's letter of December 20, 1985.
By letter dated June 20, 1986, the applicant committed to provide this in FSAR Arendment 59.
Conclusion The applicant's proposed use of. a Seabrook-specific cable tray support qualifi-cation program utilizing both test and analysis approaches in lieu of the appli-cable FSAR criteria previously reviewed and accepted by the staff is acceptable with the above applicant agreed-on commitment to provide (1) full documentation of its basis of qualification for each cable tray support in its installed con-figuration according to the program proposed in the applicant's letter of Decem-t>er 20,1985, and (2) an FSAR amendment reflecting the proposed changes listed in Attachment 2 to the applicant's letter of December 20, 1985.
I 3.9 Mechanical Systems and Components 7
k 3.9.6 Inservice Testing of Pumps and Valves By letters dated December 31, 1985, and June 4 June 6, June 18 June 23, and June 25,1986, the applicant submitted an inservice test (IST) program for pumps and valves. The applicant has stated that the IST program will meet the requirements of 10 CFR 50.55a(g), including the 1980 Edition of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code) Sec-tion XI, through Winter 1981 Addenda.
The applicant has requested relief from these code requirements pursuant to 10 CFR 50.55a(g)(5) for certain pump and valve tests.
Seabrook SSER 5 3-5
The staff has completed a preliminary review of the IST program.
This review includes an evaluation of the applicant's general IST program and of the appli-cant's request for relief from certain ASME Code requirements for certain pumps and valves. The requests for relief are based on the applicant's determination that ASME Code requirements for certain pumps or valves are impractical within the limitations of design, geometry, and accessibility.
The staff will recom-mend granting the relief where the applicant's request has satisfied the re-quirements of 10 CFR 50.55a(g)(6)(i).
The applicant's relief requests for the specific components are primarily in terms of frequency of testing and the use of alternate test methods and parameters.
On the basis of its review of the IST program, the staff concludes that the general program meets the requirements of 10 CFR 50.55a(g) for low power (5%)
plant operation. The conclusion is based on the know' ledge that the pumps and valves are new and have seen little or no wear (therefore, they are expected to be operable when called upon to function), they have been pretested to demon-strate operability, and the minimum interval for inservice testing is 90 days for some of the components.
This minimum interval is sufficient time for the staff to meet its commitment to issue a completed safety evaluation on the applicant's IST program At the time of issuance of the completed safety eval-uation, the staff will confirm the IST program is acceptable for licensing for full power operation.
If during the staff's actions to complete the safety evaluation, certain pumps and/or valves are identified that are not categorized as ASME Code, Class 1, 2, or 3, but perform a safety function, the staff will require that those pumps and/or valves be included in the IST program.
In reviewing the applicant's relief requests, due consideration has been given to the b;.rden on the applicant that could result if certain Code requirements were imposed on the facility.
These burdens included the forced shutdown of i
the plant to perform tests, performing tests on certain components on a 90-day freq1ency when considerablyiless frequent testing may be justified, activating spray systems that mav be harmful to equipment, and reducing redundancy in safety trains, sa'. ;y quipment, and safety components during plant operation ir order to b wie o perform tests imposed by the Code requirements.
Jhe stm f && U.4t interim relief from certain ASME Code requirements will nn mm ggr Hfe or property or the common defense and security of the public and a
- c. ine public interest giving due consideration to the burden on the applicant that could result if the requirements were imposed.
3.10 Seismic and Dynamic Qualification of Safety-Related Mechanical and Electrical Equipment 3.10.1 Seismic and Dynamic Qualification 3.10.1.1 Introduction Evaluation of the applicant's program for seismic and dynamic qualification of safety-related electrical and mechanical equipment consists of (1) a determina-tion of the acceptability of the procedures used, standards followed, and the completeness of the program in general, and (2) an audit of selected equipment items to develop a basis for the judgment of the completeness and adequacy of the implementation of the entire seismic and dynamic qualification program.
Seabrook SSER 5 3-6 i
Institute of Electrical and Electronics Engineers (IEEE) Standards 344-1975 and i
323-1974. These documents define acceptable methodologies for the seismic qualification of equipment.
Conformance with these criteria is sufficient to satisfy the applicable portions'of GDC 1, 2, 4,14, and 30 of Appendix A to 10 CFR 50; Appendix B to 10 CFR 50; and Appendix A to 10 CFR 100. The program is evaluated by the Seismic Qualification Review Team (SQRT), wnir.h consists of engineers from NRC and the Idaho National Engineering Laboratory (INEL, EG&G -
Idaho).
3.10.1.2 Discussion The SQRT reviewed the equipment dynamic qualification information in FSAR Sec-ticns 3.9.2 and 3.10 and visited the plant from November 5 through November 8, 1985, to determine the extent to which the equipment installed at Seabrook Unit 1 meets the criteria described above.
A representative sample of safety-related electrical and mechanical equipment as well as instrumentation, in both the nuclear steam supply system (NSSS) and the balance of plant (BOP) scopes, l
was selected for the audit.
Table 3.10-1 identifies the equipment audited.
The plant-site visit included field observations of the actual, final equipment configuration and its installation.
Observing the field installation of the equipment is necessary to verify and validate equipment modeling used in the c; qualification program.
These observations were followed by a review of the corresponding design specifications and test and/or analysis documents main-tained in the applicant's central files.
The applicant also provided details of the maintenance, startup testing, and inservice inspection.
The audit identified both generic and equipment-specific concerns.
Subsequently,"
the applicant submitted additional information resolving all of the issues.
A summary of the issues and their disposition is presented in the following sec-tions and in Table 3.10-1.
3.10.1.3 Generic Items During the field observation of,the nuclear instrumentation system cabinet, the SQRT noted that the clearance between this unit and the adjacent solid state protection system train B was not adequate.
The team also learned that this problem was associated with many other cabinets.
However, the applicant was aware of the problem and indicated that the problem was being analyzed and its resolution was being actively pursued.
In a letter dated July 16, 1986, the applicant confirmed that the generic program for preventing seismic interaction between electrical cabinets is complete, and the resolution was to modify the cabinet configuration to prevent harmful interaction caused by seismic loading.
i Therefore, this issue is closed.
During the documentation review of the reactor makeup water valve (RMW-30:
NSSS-5), it was discovered that the g-loading assumed for the valve qualifica-tion had not been reconciled with the as-built condition.
In addition, this had not been done for other valves.
However, the applicant indicated that a reconciliation program was in progress.
Subsequent documentation provided by j
the applicant confirmed the completion of this program.
Therefore, this issue j
is closed.
]
Seabrook SSER 5 3-7
The PVORT resolved all but eive of the specific operability concerns that were identified.
These five concerns follow.
(1) Auxiliary feedwater pump turbine operability with moisture in the steam was not addressed.
i I
(2) The auxiliary feedwater pump turbine end seal was cracked,'and the cause had not been determined.
1 l
(3) Operability of the auxiliary feedwater pump turbine trip and throttle valve was not ensured after an overspeed trip.
(4) Timing requirements were not addressed for control check valve FW-V-331.
I (5) Cooling tower pump SW-P-110A O-ring maintenance procedures were not addressed in accordance with the manufacturer's requirements.
In addition, the applicant was informed of five generic issues that were to be addressed before fuel load.
These five issues follow.
(1) Not all of the preservice tests required before fuel load were completed.
(2) Approximately 10% to 15% of all pumps and valves important to safety were not yet qualified and installed.
(3) The plant maintenance procedures were not complete enough for the staff to.
determine that safety related equipment will be maintained in its qualified
- state for the life of the plant.
(4) BOP valves less than 2 in. in size were not included in the Seabrook pump l
and valve operability assurance program.
I i
(5) The FSAR active valve lists were not current.
(
l These concerns and issues wer'e confirmatory and form the basis for the discussion l
l presented in Section 3.10.2.3 below.
After the site audit, the applicant submitted letters dated December 31, 1985, April 8,1986, May 1,1986, and June 13, 1986, which resolved all of the speci-fic issues and four of the generic issues.
The remaining generic issue will be I
j verified complete by the Region I staff.
The manner in which each confirmatory l
l issue was addressed is briefly discussed in Section 3.10.2.3 and is indicated in Table 3.10-3.
l The PVORT has found that the applicant is dealing with the equipment qualifica-tion issue in a positive manner. All of the SER items were adequately resolved through additional clarifications and appropriate commitments provided oy the applicant. During the audit, the applicant addressed all questions posed by the PVORT and committed to resolve all audit issues before fuel load.
Further-more, the applicant discussed significant aspects of the overall equipment qua-lification program-such as amplified response spectra reconciliation, equipment modification and reconciliation of original qualification reports, nozzle load Seabrook SSER 5 3-10
l The applicant was to provide written confirmation in th2 FSAR that all i
active BOP valves are covered by the Seabrook pump and v (2) valves smaller than 2 in, have been included in its valve operability assurance program.
E qualification program.
The appropriate sections of the Seabrook FSAR have Applicant Response:
1 been revised by Amendment 56, resolving this generic issue.
it was apparent that a complete list At the conclusion of the PVORT audit, The applicant was to of active valves had not been provided in the FSAR. confir (3)
The safety-related BOP and NSSS valves have been iden-This tified in FSAR Tables 3.9 (8)-25 and 3.9 (N)-11 by Amendment 56.
Applicant Response:
issue is resolved.
At the time of the audit, most construction tests had already been com-The However, the hot functional tests were still in progress.
(4) applicant was to confirm that all preservice tests that are required be-I pleted.
fore fuel load have been completed.
I Initially, the applicant committed to complete preser-I Applicant Response:
Subsequently, in Supplement No. 4, the vice tests before fuel loading. staff noted that, in a letter dated April E
P i
mitted to complete the preservice testing before commercial operat In a letter dated June 24, 1986, before the component is required to be operable fo in The staff finds that although this repre-i the Technical Specifications.
sents a considerable improvement in the schedule for perfonning and com-pleting the preservice testing, further discussions will be needed befor ll the staf f will consider a modification of the requirement to complete a preservice tests before fuel loading.
l accepted by the staf f, this remains a confirmatory item.
f At the time of the audit, approximately 10% to 15% of all pumps and valve The applicant shall confirm (5) important to safety had not been qualified.
d that all pumps and valves import nt to safety are proper that the original loads used in tests or analyses to qualify pumps and installed.
h valves important to safety are not exceeded by any new loads, such as imposed by a loss-of-coolant accident (LOCA) (hydrodynamic loads) or built conditions.
In a letter dated July 1, 1986, the applicant stated d
that all pumps and valves requiring qualification are now qualif Applicant Response:
are installed in the plant.
On the as-built conditions have been reconciled to the origi 3-13 Seabrook SSER 5 9
9
1 1
s 4
REACTOR 4.4' Thermal-Hydraulic Design 4.4.5 Instrumentation
{
4.4.5.3 Loose Parts Monitoring Systems In the SER, the staff stated that the applicant's loose parts monitoring system (LPMS) was similar to the LPMS at the Shoreham nuclear plant, which was approved by the HRC staff. The applicant committed to provide a detailed equipment de-scription demonstrating compliance with RG 1.133.
This was submitted in Amend-sents 55 and 56 to the FSAR anet ;,covided information including the following.
l T. e LPMS will consist of 12 active instrumentation channels, each comprising a 5
piezoelectric acceleror.eter (sensor), signal conditioning equipment, and diag-nostic equipment. Two redundant sensors are fastened mechanically to the reac-i tor coolant system at each of the following potential loose parts collection l
regions:
{
f (1) reactor pressure vessel--upper head regions (2) reactor pressure vessel--lower head region l
l (3) each steam generator-reactor coolant inlet region The system will be capable of detecting a metallic loose part that weighs from 0.25 lb to 30 lb impacting in the vicinity of six natural collection regions having a kinetic energy of 0.5 f t-lb on the surf ace.
The system is designed to remain functional for a seismic event up to and in-cluding the operating-basis earthquake.
The staff has found that the applicant's submittal describing compliance of the l
LPMS with RG 1.133 is acceptable because the applicant has satisfactorily ad-J dressed the specific positions of Section C and has not requested any deviations.
5ection C.3 of RG 1.133 states that alert levels for startup and power opera-tion have to be established and must be submitted to the NRC staff in a final baseline report within 90 days following completion of the startup test program.
The applicant has committed (letter from J. DeVincentis, PSNH, to V. Noonan, j
NRC, dated April 8,1986) to provide, before power operation, a final base-j line report that will contain (1) an evaluation of the LPMS for conformance to RG 1.133, " Loose-Part Detection Program for the Primary System of Light-Water-Cooled Reactors"; (2) a description of the system hardware, operation, and im-plementation of the loose parts detection program, including the plans for startup testing, acquisition of baseline data, and alarm settings; and (3) a description and evaluation of diagnostic procedures used to confirm the pre-sence of a loose part. The applicant has stated in FSAR Amendment 55 that a training program will be provided to give pertinent general and specific l
1 l
Seabrook SSER 5 4-1
training for plant personnel involved with system operation, maintenance, and loose part diagnosis before station operation.
This is acceptable to the staff.
,l Cn the basis of the applicant's conformance with RG 1.133 including its com-I sitzent to supply the baseline report following startup testing, the staff I
l concludes that the LPMS for Seabrook is acceptable.
- 4. 4. 5. 4 ICC Instrumentation On November 4,1982, the Commission determined that instrumentation system for detection of inadequate core cooling (ICC) consisting of upgraded subcooling s.argin monitor, core exit thermocouple, and a reactor coolant inventory track-i t.;; system is required for the operation of pressurized-water-reactor facili-ties.
The requirements of the instrumentation system for the detection of ICC are delineated by NUREG-0737, Item II.F.2.
The applicant responded by submit-ting design information, revised FSAR pages, emergency operating procedure guidance, and design analysis and test d6ta in a letter dated February 24, 1986.
J Technical Specifications covering ICC are undergoing a separate review as part i
of the overall Seabrook Technical Specification review. The applicant states that emergency operating procedures when written will conform to approved emer-gency response guidelines.
A separate review of emergency operating procedures is conducted only if the applicant deviates from the approved guidelines.
The Seabrook ICC monitoring system will be installed and tested before fuel is loaded and will be calibrated before the plant achieves 5% power.
The.1C0 monitoring system installed at Seabrook Station includes (1) core exit thermocouple monitoring (2) core subcooling margin monitoring (3) reactor vessel level monitoring The incore thermocouple / core cooling monitor has been installed to provide im-proved information presentation and display to the plant operators on the status of core heat removal capability.
The system monitors all core exit thermocouple and calculates core subcooling margin utilizing redundant channels of instrumen-tation and control room displayh.
The core exit temperature monitoring system is a part of the incere instruments-tion system, which consists of 58 thimble assemblies, each with 5 fixed neutron detectors and 1 Type K chromel-alumel thermocouple at fixed core outlet posi-tions. The core exit thermocouple monitoring system uses two redundant, inde-pendent trains that monitor all 58 of the Seabrook Station chromel-alumel core i
exit thermocouple (29 on train A and 29 on train B).
The analog indicator dis-plays the third highest valid thermocouple temperature. The train and quadrant orientation of the thermocouple were provided by the applicant. The core exit thermocouple leads leave the core through the bottom of the reactor vessel.
Af ter leaving the vessel, they are routed for mating with the extension wires (cables) through the seal table where the thermocouple leads are teminated with qualified connectors. The applicant stated that, beyond the seal table, the cables are routed to the containment penetrations in a manner consistent with RG 1.75. The uncompensated core exit thermocouple signals (29) for each train are then sent to their respective ICC monitors, which house the reference junctions. The thermocouple have a range of 0* to 2300*F.
The thermocouple Seabrook SSER 5 4-2 l
1
]
information is provided as an output on two redundant plasma displays, which se-ve as both the primary and backup displays.
ir The subcooling mar' gin monitor calculates the margin to saturation temperature us 4 ng wide-range reactor coolant system pressure and a core exit temperature value based on the auctioneered high thermocouple quadrant average temperatures.
This represents bulk loop temperature and is consistent with the Westinghouse The subcooling margin calculated 0.ners Group Emergency Response Guidelines.
The values are routed to redundant plasma displays and analog indicators.
a applicant stated that the cable routing from sensor input to display meets RG 1.75.
~
The upper and lower range limits of the core subcooling calculation are adjust-It was shown that able up to a reactor coolant system pressure of 3000 psia.
the RG 1.97 range requirements of 200*F subcooling and 35'F superheat are i
I satisfied.
The monitoring system displayr several levels of information including (1) bulk average core exit thermocouple trending, (2) a spatial map exhibiting the ther-ooccuple temperature at its respective location in the core, (3) a core map showing minimum, average, and maximum quadrant temperatures, (4) subcooling mar gin, (5) a detailed data list exhibiting thermocouple location, tag designation, and temperature, and (6) hot channel core exit temperature.
i The reactor vessel level instrumentation system (RVLIS) consists of two redun-Each dant, independent trains that monitor the reactor vessel water levels.
full range and dynamic head.
The.
train provides two vessel level indications:f ull-range RVLIS reading p f rom the bottom of the vessel to the top of the vessel during natural circula-The dynamic head reading provides an indication of reactor tion conditions.
core, internals, and outlet nozzle pressure drop for any combination of operat-l Comparison of the measured pressure drop with the ing reactor coolant pumps.
normal, single-phase pressure drop provides an approximate indication of the i
relative void content of the circulating fluid.
Two analog indicators (per train) in the control room display the RVLIS read-The applicant stated that ings, one for dynamic head and one for full range.
The range of the the cable routing from sensor input to display meets RG 1.75.
full-range channels corresponds to 0% to 120% of the reactor vessel height.
The inside bottom of the vessel to the inside top of the vessel head is approx-This is equivalent to 100% of the full-range level.
imately 41 f t.
Evaluation and Conclusions 24, 1986, and The staff has reviewed the applicant's submittal dated February its conclusions are:
The commitment to complete installation and testing before fuel is loaded and calibration before the plant achieves 5% power is acceptable.
(1)
For new fuel the quantity of decay products in the core is extremely low; hence, the residual heat in the core following a shutdown is also low and There-This lowers the risk of inadequate core cooling.
easily removed.
fore, the health and safety of the public is not compromised as long as 4-3 Seabrook SSER 5
calibration is completed and the systems are fully operational before 5%
power is exceeded.
(2) The implementation schedule for the ICC instrumentation system is not de-scribed.
However, the staff will require that the ICC instrumentation sys-tem shall be fully operational with appropriate emergency operating proce-dures in place before fuel is loaded and that the system 56all be fully calibrated before 5% power operation is exceeded.
The acceptability of the final design will be demonstrated by the successful completion of preopera-tional testing of the ICC. subsystems.
An implementation letter report is required to complete the staff's review for implementation approval of the installed ICC system and must be provided before 5% power operation is ex-ceeded. The implementation letter report should contain (a) notification that the system installation, functional testing, and calibration are complete and test results are available for inspection (b) summary of applicant conclusions based on test results, e.g.,
performance of the system in accordance with design expectations and within design error tolerances, or description of deviations from design performance specifications and basis for concluding that the deviations are acceptable (c) description of any deviations of the as-built system from previous design descriptions with any appropriate explanation (d) confirmation that the emergency operating procedures (EOPs) used for operator training are complete and conform to the technical content of HRC-approved E0P guidelines (generic or plant specific)
(3) The core exit temperature monitors, subcooling margin monitors, and reac-tor vessel level monitors meet the single-failure criteria of RG 1.53.
No single failure within the ICC instrumentation or its auxiliary support equipment will render the system inoperable.
(4) FSAR Section 7.5.4.4 states that ICC variables are all Category 1 and the ICC equipment is environmentally qualified in accordance with RG 1.89.
The quality assurance of the ICC Category 1 instrumentation will be in ac-cordance with the QA program described in FSAR Chapter 17.
This is con-sidered provisionally acceptable and may be confirmed by inspection at a future date.
On the basis of its review of FSAR Sections 4.4.6.5 and 7.5 and the additional information provided by the applicant, the staff finds that the design of the l
Seabrook Station ICC instrumentation system is in conformance with NUREG-0737, i'
Item II.F.2, and is, therefore, acceptable.
4.4.8 Sunnary and Conclusions In the SER, references to NUREG-0737. Item II.F.2, " Instrumentation for Detec-tion of Inadequate Core Cooling," were made in this section. This subject matter is fully addressed in Section 4.4.5.4 of this supplement.
i Seabrook SSER 5 4-4
f and natural circulation tests to be conducted at Diablo Canyon to the Seabrook design." The Diablo Canyon boron mixing and natural circulation test was con-I ducted on March 28 and 29,1985, at Unit 1, and is reported in Westinghouse Topical Reports WCAP-11086 (proprietary) and WCAP-11095 (nonproprietary).
By letter dated June 1,1986, the applicant submitted a report (YAEC 1552) de-scri,bing the applicability of the Diablo Canyon Unit I natural circulation and boron mixing test to Seabrook.
Analyses of natural circulation in a pressurized-water reactor (PWR) and scaled experiments conducted under Electric Power Research Institute (EPRI) contract show that transient behavior in natural circulation is governed by heat trans-fer in the steam generator, while, for steady state, the important steady-state parameters (flowrate and core temperature difference) depend on flow resistance (EPRI Report NP-1363-SR). Diablo Canyon Unit I uses four 51 series steam genera-
- t. ors, and Seabrook will use four model F steam generators.
Although the designs are not identical, the steam generators have approximately the same heat-transfer-area-to power ratios, and the elevations of the U tubes above the top of the fuel assemblies are approximately the same.
The differences in flow resistance in the reactor coolant systems (RCSS) of the two plant designs are not significant for this natural circulation capability comparison.
The applicant identified five areas in which a shutdown from full power and cooldown to cold shutdown would differ from the test performed at Diablo Canyon Unit 1.
These are:
(1) method of RCS depressurization j
J (2) upper head temperature (3) boron source (4) residual heat removal initiation pressure
(.5) reactor vessel internals The staff has reviewed these differences between Diablo Canyon and Seabrook, which could be significant with' respect to natural circulation cooldown, and finds that for these differences the comparisons are favorable.
However, the i
staff is still reviewing the Diablo Canyon test report (WCAP-11086) and is also performing its own studies of what design and operation considerations should be considered in a comparison of the Diablo Canyon test to a given Westinghouse-designed PWR. Therefore, although the applicant has satisfied the staff require-ment by providing a report describing the applicability of the Diablo Canyon Unit I test results to Seabrook, the staff considers this item to be confirma-tory pending completion of the staff review of the Diablo Canyon test report (WCAP-11086).
Should any further questions arise as a result of the review of WCAP-11086 that are applicable to Seabrook, the applicant will be expected to provide the neces-sary responses. In a letter dated July 18, 1986, the applicant committed te provide any additional infonnation that may be required in response to questions arising from the staff's further review.
Operation of Seabrook before the completion of the staff's review of WCAP-11086 is based on the following. The applicant has successfully performed a loss-of-offsite power test during startup testing. Although this test does not Seabrook SSER 5 5-3
i demonstrate the ability to reach cold shutdown using only safety-related equip-l ment, it does demonstrate that stable natural circulation cooling can be j
achieved.
In addition, Westinghouse-designed reactors have successfully cooled down using natural circulation (Lanning and Wunderlick, 1984).
Accordingly, the staff finds operation of Seabrook before completion of the staff's review of WCAP-11086 to be acceptable.
l 1
l l
I l
I I
I l
1
\\
\\
I l
1 l
l Seabrook SSER 5 5-4 l
1
l' 6.2.6 Containment Leakage Testing Program Containment Air Lock Surveillance By letter dated May 21, 1986, the applicant requested an exemption from certain requirements of Appendix J to 10 CFR 50.
{
Paragraph III.D.2(b)(ii) of Appendix J states:
" Air locks opened during periods when containment integrity is not required by the plant's Technical Specifications shall be tested at the end of such periods at not less than Pa."
Whenever the plant is in cold shutdown (mode 5) or refueling (mode 6), con-tainment integrity is not required.
However, if an air lock is opened during modes 5 and 6, Paragraph III.D.2(b)(ii) of Appendix J requires that an overall air lock leakage test at not less than Pa be conducted before plant heatup and
{
startup (i.e., entering mode 4).
The existing air lock doors are so designed that a full pressure (i.e., Pa (49.6 psig)) test of an entire air lock can only
{
be performed af ter strong backs (structural bracing) have been installed on the inner door.
Strong backs are needed because the pressure exerted on the inner i
j door during the test is in a direction opposite to that of the accident pressure direction.
Installing strong backs, performing the test, and removing the strong backs requires several hours, during which access through the air lock is prohibited.
{
When no maintenance has been performed on the air lock that could affect its sealing capability, and the air lock doors have been closed in accordance with j
the applicant's procedure, and the periodic 6-month test at Pa required by Para-graph III.D.2(b)(i) of Appendix J has been performed on schedule, there is no reason to expect the air lock to leak excessively just because it has been opened in a shutdown or refueling mode.
Performing the door seal leak test of Paragraph III.D.2.(b)(iii) of Appendix J is sufficient, in this case, to demon-strate the continuing integrity of the air lock.
Accordingly, the staff concludes that the applicant's proposed approach of sub-stituting the seal leakage test of Paragraph III.D.2(b)(iii) for the full-pressure test of Paragraph III.D.2(b)(ii maintenance that could affect sealing cap) ability has been performed on an a lock.
Whenever maintenance that could af fect sealing capability has been per-formed on an air lock, the requirements of Paragraph III.D.2(b)(ii) of Appen-dix J must still be met by the applicant.
The special circumstances for granting this ' exemption pursuant to 10 CFR 50.12 have also been identified.
The purpose of Appendix J to 10 CFR 50 is to ensure that containment leaktight integrity can be verified periodically throughout service lifetime so as to maintain containment leakage within the limits speci-fied in the facility Technical Specifications.
The proposed alternative test method is sufficient to achieve this underlying purpose in that it provides adequate assurance of continued leaktight integrity of the air lock.
In addi-tion, at the time this section of Appendix J was revised in 1980, the staff did not contemplate the undue hardship and cost that would result from the requirement to perform a time-consuming full pressure test before starting up from even the shortest cold shutdown during which the air lock had been used Seabrook SSER 5 6-3
f i
appropriate steps are taken by GA Technologies to make sure that use of the f
firmware configuration document will result in use of the correct version /
revision of a PROM.
By letter dated June 26, 1986, the applicant has committed to do so.
On the basis of the above discussion, the staf f considers the V&V procedures used during the design and testing of the Seabrook Class IE RDMS adequate.
This issue is, therefore, considered resolved.
7.5.2.4 Postaccident Monitoring Instrumentation The applicant was requested by Generic Letter 82-33 to provide a report to the NRC describing how the postaccident monitoring instrumentation meets the guide-lines of RG 1.97 as applied to emergency response facilities.
The applicant's response to RG 1.97 was provided by letter dated August 30, 1985, and Revi-sion 58 of the FSAR.
EG&G Idaho, Inc., under contract to the NRC, with general supervision by the NRC staff, performed a detailed review and technical evaluation of the appli-cant's submittals.
This work was reported by EG&G in the Technical Evaluation Report (TER), "Conformance to Regulatory Guide 1.97, Seabrook Station, Unit Nos. I and 2," dated May 1986 (attached as Appendix L to this supplement).
The staff has reviewed this report and concurs with the conclusion that the appli-cant either conforms to, or has adequately justified deviations from, the guidance of RG 1.97 for each postaccident monitoring variable except for accu-mulator tank level and pressure, quench tank temperature, and containment sump water temperature.
Af ter the generic letter was issued, the NRC staff held regional meetings in February and March 1993 to answer licensee and applicant questions and concerns regarding the NRC policy on RG 1.97.
At these meetings, it was established that i
the NRC review would only address exceptions taken to the guidance of RG 1.97.
I Furthermore, if licensees or applicants explicitly state that instrument systems conform to the provisions of the regulatory guide, no staff review would be necessary for those items.
Therefore, the review performed and reported by EG&G only addresses exceptions to the guidance of RG 1.97.
This evaluation addresses the applicant's submittals on the basis of the review policy described in the NRC regional meetings and the conclusions of the review as reported by EG&G.
The staff has reviewed the evaluation performed by EG&G contained in the TER and concurs with its bases and findings.
The applicant either conforms to, or has acceptably justified deviations from, the guidance of RG 1.97 for each postaccident monitoring variable except accumulator tank level and pressure, quench tank temperature, and containment sump water temperature.
For the variable, effluent radioactivity--noble gases, addressed in Sec-tion 3.3.5 of the EG&G TER--the applicant, by letter dated July 18, 1986, pro-vided markert-up copies of FSAR Table 7.5-7, Sheet 12, and Page 7A-9, which reflects the change of this variable from Design Category 2 as specified in RG 1.97 to Design Category 3.
The applicant's justification for this change is that the main plant vent stack noble gas radiation monitor, which is Design Category 2, provides the primary indication for detection of significant re-leases on breach of the containment.
The containment enclosure noble gas Seabrook SSER 5 7-12 e
4 radiation monitor (the instrument that measures the subject variable) is backup indication that assists the operators in identifying the location of a release that has been identified by the main plant vent stack monitor.
Since backup Type C variables are considered Design Category 3, the applicant is revising the FSAR to reflect this change and will not environmentally qualify this instrument.
Because the applicant is using this instrument as backup to an environmentally qualified Design Category 2 instrument, the staff agrees,.
I that it may be classified as Design Category 3 and need not be environmentally qualified.
For the variable, boric acid charging flow, addressed in Section 3.3.25 of the EG&G TER, Revision 58 of the Seabrook FSAR provided a justification for its classification as a Design Category 3 variable rather than as Design Category 2 as specified in RG 1.97.
This justification was not available to EG&G at the time it was conducting its review of this item.
The justification states that emergency boration is not required in the mitigation of design-basis accidents.
The refueling water storage tank provides the required volume of borated water for all design-basis accidents.
Emergency boration may. be used to assist in the recovery, if available.
Therefore, the monitoring of emergency boration flow is classified as Design Category 3.
Because eraergency boration flow is not a safety injection flow at this station, the staff finds that classifica-tion of this variable as Design Category 3 is acceptable.
With regard to pressurizer heater status addressed in Section 3.3.10 of the EG&G TER, the applicant, in a letter dated May 30, 1986, provided a revision l
l to Table 7.5-1, which now indicates that this variable is categorized as Design,
Category 2 and is, therefore, environmentally qualified.
This is in accordance I
with the recommendations of RG 1.97 and is, therefore, acceptable.
With regard to accumulator tank level and pressure addressed in Section 3.3.7 of the EG&G TER, the applicant, by a letter dated June 6,1986, has committed to install an environmentally qualified accumulator tank level or pressure instrument that meets the requirements for a Category 2 instrument contained in RG 1.97.
In the letter, the applicant also has committed to install an envi-ronmentally qualified containment sump water temperature instrument that meets the Category 2 RG 1.97 requirements.
Containment sump water temperature is addressed in Section 3.3.17 of the EG&G report.
The applicant has committed to install these instruments before startup following the first refueling outage. The staff finds these commitments acceptable.
With regard to the quench tank (pressurizer relief tank (PRT)) temperature addressed in Section 3.3.11 of the EG&G TER, the applicant i.s supplying a temperature range of 50'-250'F, which deviates from the recommended range of 50*-750*F.
The applicant's justification is that the PRT temperature measure-ment is used in conjunction with PRT level and pressure indication to determine if PRT conditions are normal.
The applicant states that the temperature range provided is sufficient to irdicate a change from normal operating conditions, and fully enveloping the expected temperature to be encountered is not neces-sary in this case. The staff finds that the applicant's justification is not acceptable. The applicant should show that the indication of temperature, including the maximum expected saturation temperature, will remain functional and on scale during any accident that lifts the pressurizer relief valves, or provide a range that will envelop these conditions.
{
Seabrook SSER 5 7-13 L_-___-
Criterion (5) l The Seabrook service water system uses seawater for cooling water, but the de-sign provides a double barrier between the primary coolant system and the sea-f Grab sample analysis for chloride on diluted primary coolant liquid water.
samples will be completed within 96 hours0.00111 days <br />0.0267 hours <br />1.587302e-4 weeks <br />3.6528e-5 months <br /> of drawing the sample'. Chloride con-centrations in the liquid samples are determined by ion chromatography, which t
V has a detection limit of 10 parts per billion.
Therefore, the sample can be diluted by as much as a factor of 15 and still provide the required minimum detectable threshold for chloride.
I The PASS can dilute samples by as much as a fe.cor of 1000 for as-low-as-is-I reasonably-achievable (ALARA) purposes, which would result in a chloride detec-j i
tion threshold of 10 parts per million.
The PASS can also take an undiluted sampleJhich would be retained for up to 30 days before analysis 6e prev 4de for short-lived radioactivity decay, consistent with ALARA considerations.
e
)
The staff concludes that these provisions meet Criterion (5) in Item II.B.3 I
of NUREG-0737 and are, therefore, acceptable.
Criterion (6)
The applicant has performed a shielding analysis to ensure that radiation ex-posure from reactor coolant and containment atmosphere sampling and analysis activities is within the acceptable limits of 5 rem whole body and 75 rem ex-tremities to the operator.
The operator exposure will include entering and leaving the sample panel area, operating the sample panel manual valves, per-l forming manual sample dilutions, and transferring samples to the chemistry laboratory for analysis.
The staf f has determined that these provisions meet Criterion (6) in Item II.B.3 of NUREG-0737 and are, therefore, acceptable.
Criterion (10)
{
The accuracies, ranges, and sensitivities of the PASS instruments and analyt-ical procedures are consistent with the recommendations of RG 1.97, Revision 3, and the clarifications of NUREG-0737, Item II.B.3.
Therefore, they are ade-quate for assessing the radiological and chemical status of the reactor coolant.
The instrumentation was selected for its ability to operate in the postaccident j
sampling environment.
Equipment used for postaccident sampling and analysis i
will be calibrated or tested at least once every 6 months.
Retraining of operators for postaccident sampling is scheduled at a frequency of once every 6 months.
The staff finds that these provisions meet Criterion (10) in l
Item II.B.3 of NUREG-0737 and are, therefore, acceptable.
Conclusion On the basis of the above evaluation, the staff concludes that the applicant's proposed PASS meets all 11 criteria in Item II.B.3 of NUREG-0737 and is, there-fore, acceptable.
The previously proposed license cr>ndition on the PASS can j
now be removed.
i l
i Seabrook SSER 5 9-3 I
J
and the structural integrity of the remaining structure was analyzed.
Where it was determined that the structural integrity of the fire area would not be com-promised by failure of the structural elements, the steel was not protected.
Those structural steel members that were determined to be required to ensure the structural integrity of the area have been protected in accordance with Section C.S.b of BTP CMEB 9.5-1.
On the basis of both this evaluation and the aforementioned Limerick evaluation,
)
the staff concludes that the absence of fire protection for the structural steel elements identified in the applicant's April 24, 1986 letter is an acceptable deviation from Section C.S.b of BTP CHEB 9.5-1.
In Supplement No. 4, the staff expressed a concern that fire-induced high-impedance faults could result in the loss of the necessary power, multiple supply 4
for safe shutdown equipment.
By letter dated May 13, 1986, the applicant pro-i vided additional information.
Multiple high-impedance faults can occur when seversi cables from a common bus are located in the same fire area.
When a fire occurs in such an area, the re-sulting fire damage could cause electrical faults in the cables, but the faults 3
i may not trip the individual circuit breakers.
However, the sum of the faults may be sufficient to trip the main breaker, which protects the power supply bus.
i I
If safe shutdown equipment is energized from the same bus, once the main breaker l
trips, this equipment will have lost its power source. The applicant stated that, for such a condition to occur, each branch circuit cable would have to
,d fail along its length so as to result in a unique insulation resistance.
This unique resistance would produce a leakage current to ground or conductor to con-ductor which, when added to the conductor load, would result in a current just below the protective device rating.
Thus, an improbable combination of tempor-ary leakage currents would have to occur at once to produce a trip of the main l
breaker.
The applicant committed to add a general note to the safe shutdown and cooldown procedures to alert plant personnel to clear nonessential loads from a tripped bus before reclosing the bus feeder breaker, should the bus feeder breaker trip.
The staff finds the applicant's provision of the note to drop the nonessential loads before reclosing a tripped breaker acceptable in that it resolves the problem of a breaker trip because of high-impedance faults.
Ventilation In the SER, the staff reported that charcoal' filters have been provided with low-flow air bleed systems in accordance with RG 1.52.
During the audit of the applicant's fire protection program for Seabrook Unit 1, the staff obs ved that only those charcoal filters that have been designated as enginee fety X
features (ESFs) have air-bleed systems.
By letter dated March 18, 198, the applicant r sted a deviation from installing fire suppression systems to protect c a coal filters including both ESF and non-ESF filters. The appli-cant indicated that an analysis was conducted which showed that the temperature for adsorber autoignition or desorption would not occur even without air flow across the adsorbers. Previously, the Safety Analysis Report indicated that a low-flow air-bleed velocity of 0.2 f t per minute was required to maintain the Seabrook SSER S 9-8
adsorber temperature below the autoignition or desorption point.
Staff discus-sions with the applicant concerning its analysis indicate that certain assump-tions used in the analysis require further justification.
l Staff review of this issue is still in progress.
The resolution of this issue will be provided in a future supplement to the SER befcre an operating license is issued.
This is an open item.
Lighting and Communication In the SER, the staff reported that the applicant committed to install 8-hour battery powered emergency lighting units in all areas of the plant that may J
need to be staffed for safe shutdown merations, and in access and egress
)
routes to and from all such areas, in accordance with Section C.S.g of BTP
?
CHEB 9.5-1.
l i
By letter dated January 24, 1986, the applicant requested deviations from Sec-tion C.S.g of BTP CMEB 9.5-1 in the control room, switchgear room A, and switch-I gear room B.
In Supplement No. 4, the staff reported that its review of these deviation requests was ongoing.
Following is the staff's evaluation of these deviations.
In the control room (fire area CB-F-3A-A) diesel generator powered lighting is provided to illuminate the face of the main control board and the control switches used to achieve safe shutdown.
This lighting is powered from both diesel gen-erators in an "A and/or B" logic so that lighting is provided by either train.
l The diesel generator engine rooms are separated from each other by 3-hour-fire,
rated barriers. Power from each diesel generator is routed to separate switch-gear rooms (A and B), which are separated from each other and from the diesel generator rooms by 3-hour-fire-rated barriers.
Power from each switchgear room is routed via embedded conduit on opposite sides of the control building to f
train A and train B essential lighting panels within the control room complex
)
computer room.
Therefore, the staff finds at least one train of power will be l
l available for control room emergency lighting for any postulated fire outside the control room complex.
In the event of a computer room fire, both essential lighting panels could be damaged.
However, in the event of such a fire, the control room will be evacu-ated and the remote shutdown system (RSS) control panels will be used to achieve shutdown. Emergency lighting for access and egress routes from the control room to the RSS facilities is provided by B-hour battery powered lights in accordance with Section C.5.g of BTP CMEB 9.5-1.
Eight-hour battery powered emergency lighting units are provided in switchgear room A for all RSS control locations and access / egress routes thereto, with the exception of EDE-MCC-512, EDE-MCC-515, and EDE-MCC-521. These panels (motor control centers (HCCs)) are illuminated by fluorescent light fixtures powered by the train B diesel generator.
The power feeds to these light fixtures are confined to switchgear rooms A and B, which are separated from each other and from each diesel generator room by 3-hour-fire-rated barriers. Therefore, the lighting in switchgear room A will not be disabled by a fire outside the switch-gear rooms and the diesel generator rooms. Moreover, access to these MCCs is not required for a fire in either switchgear room or diesel generator room.
Seabrook SSER 5 9-9
safety-related components in accordance with the staff fire protection guide-lines as reported in the SER and its supplements, the staff does not expect this delay to adversely impact the level of fire safety.
The use of portable l
fire extinguishers and the substitution of fire hoses from the fire hydrant hose houses for standpipe system hose is, therefore, an acceptable deviation from Section C.6.c of BTP CMEB 9.5-1.
~
9.5.1.8 Summary of Approved Deviations From BTP CMEB 9.5-1 I
The following deviations from BTP CMEB 9.5-1 were approved in the SER and Sup-plement No. 4 (SSER 4):
ccrpet in the control room (SER and SSER 4) lack of an automatic fire suppression system and 20-ft separation between redundant safety-related equipment required for safe shutdown in certain fire areas (SER and SSER 4) drains in the switchgear rooms (SER) 1500 gal diesel fuel day tanks (SER) lack of independence for the alternate shutdown capability from certain fire areas (SSER 4) non-fire-rated wall (SSER 4) lack of fire dampers in certain heating, ventilation, and air conditioning ducts (SSER4) non-U1.-labeled dampers in certain fire areas (SSER 4) lack of areawide fire detection in certain fire areas (SER and SSER 4 fire protection water supply tanks' capacity less than 300,000 gal (SSER 4) non-UL-listed water supply valves (SSER 4)
On the basis of the above evaluation, the staff concludes that the following additional deviations are acceptable.
The section they are discussed are enclosed in parentheses.
of this supplement in which deviations from NFPA standards (9.5.1.1) non-fire-rated bus duct penetrations (9.5.1.4) unprotected structural steel members in certain fire areas (9.5.1.4) lack of 8-hour battery powered emergency lighting units in the control room and certain locations in switchgear rooms A and B (9.5.1.4) lack o'f smoke detectors above the control room suspended ceiling (9.5.1.5)
Seabrook SSER 5 9-12
i deviation of certain details of the SSE fire protection booster pump from NFPA 20 (9.5.I.5) use of fire hydrant hose houses in lieu of standpipe hose houses for certain fire areas (9.5.1.5) l l
- 9. 5,1. 9 Conclusions On the basis of its review, the staff concludes that, with the exception of the protection of the charcoal filter units, the applicant's fire protection pro-gram for Seabrook Station, with approved deviations, meets the staf f fire pro-tection guidelines of BTP CMEB 9.5-1 and satisfies GDC 3.
The staff will condition the operating license to require that the applicant implement and maintain in effect all provisions of the approved fire protection program.
9.5.8 Emergency Diesel Engine Combustion Air Intake and Exhaust System The SER states:
In a letter dated November 12, 1982, the applicant addressed the tornado-missile protection and protection from natural phenomena for the engine exhaust system.
The applicant stated that a deflector plate will be mounted on the exhaust stack to minimize i
the amount of precipitation (snow, ice, rain, etc.) that could enter and accumulate in the exhaust stack.
In addition, a pressure relief device will be provided to ensure diesel generator operation in the event that a tornado missile strikes the exhaust stack, Pending submittal of drawin the proposed modifications,gs showing the details and location of the staff finds the design modifi-cations acceptable.
l i
In Amendment 58 to the FSAR, the applicant modified the design of the diesel generator (DG) exhaust stacks. *The new design reduces the height of an exhaust stack to 4 ft above the DG building roof and provides an exhaust hood over the exhaust stack.
The staff has reviewed the desi and finds that it will prevent precipitation (snow, ice,gn of the exhaust hood rain, etc.) from entering and accumulating in the exhaust stack.
However, reducing the DG exhaust stack height to 4 ft introduced a new concern, which is a potential blockage of the stack resulting from drifting snow.
dated May 29, 1986, The applicant, in a letter DG exhaust stack clear of snow during a storm.has described the procedures
/
information provided and finds it acceptable.
The staff has evaluated the ing the months of November through March, the exhaust stacks will be check f
periodically af ter any significant snowfall for driftin necessary to maintain the DG in an operable condition. g snow and cleared as 1
Amendment 58 to the FSAR also states that the DG exhaust stacks are not to missile protected, but that on the basis of a probability risk analysis, the probability of a tornado damaging the exhaust stacks is low.
By letters dated May 29 and June 4,1986, the applicant provided additional information in response to the staff's concern.
The applicant has stated that an analysis was performed to estimate tornado missile impact probabilities on the as-built Seabrook SSER $
9-13 i
1
Dry active wastes will be processed in a 100-f t3 box compactor with antisprin back devices.
The compactor exhaust system operates to prevent dust from es g-caping during trash handling, and will be complete with fan, fan motor, pre-filter, absolute filter, and connections to the duct in the filled drum storage area exhaust system.
The applicant will have onsite storage capacity for 450 to 650 ' drums of solidi-fied waste.
Solid waste containers, shipping cask, and methods of packing are designed to meet applicable State and Federal regulations, including 10 CFR 71.
11.4.2 Evaluation and Findings The staff has reviewed the SWPS in accordance with the acceptance criteria of SRP Section 11.4 (NUREG-0800).
The scope of the review included line diagrams of the system, piping and instrumentation diagrams (P& ids), and the system description of the SWPS and those auxiliary supporting systems that are essen-tial to the operation of the SVPS.
The applicant's proposed design criteria and design bases for the SWPS and the applicant's analysis of those criteria and bases have also been reviewed.
The staff has also reviewed (1) the capability of the proposed system to process the types and volumes of wastes expected during normal operation and anticipated operational occurrences in accordance with GDC 60 and (2) provision for the handling of wastes relative to the requirements of 10 CFR 20 and 71 and applicable U.S. Department of Trans-portation regulations.
sI The staff has reviewed the mobile solidification system that will be provided by the NUS Corporation.
This system was approved for reference by licensees in-an NRC safety evaluation (Thomas, May 1985).
This evaluation stated that utili-zation of this system at a facility required that certain plant-specific infor-s.ation be supplied.
The applicant did this in a letter dated May 29, 1986.
By letter dated June 17, 1986, spill control methods, including type and location.the applicant also provided a descrip The staff finds this acceptable.
In addition, the mobile system waste processing will be limited to the waste processing building as noted in the May 29, 1986, submittal, and the waste classification requirements of 10 CFR 61 will be met.
The asphalt system process control program (PCP) has not been approved.
PCP has not addressed The (1) methods to verify that the waste has been solidified (2) methods for reprocessing the wastes when they are shown to be unsolidified The PCP must be approved before the processing of waste utilizing asphalt.
The staff finds that both the mobile and inplant waste processing systems con-form with the remaining acceptance criteria of SRP Section 11.4.
However, on the basis of this review, the staff has concluded that the capability of Sea-brook to store processed radwaste on site is contingent on the use of the in-plant asphalt solidification system, the crystallizer, and the degree to which various evaporators are utilized to process liquid waste. The staff has con-cerns about whether adequate stordge capacity currently exists for the volume of radwaste that would be produced by use of the mobile systems alone.
The Seabrook SSER 5 11-2
l issue of onsite storage capacity is currently being addressed by the applicant.
The results of further staff analysis will be presented in a future supplement to the SER before an operating license is issued.
This is an open item.
11.5 Procc:s_s,, and Ef fluent Radiological Monitoring and Sampling Systems 11.5.2 Evaluation Findings The staff questioned the iack of provision for sampling or monitbring the turbine gland steam condenser exhaust.
The applicant has committed to include provisions for continuously sampling this effluent stream for radiciodines and particulate. There is no noble gas monitor because the noble gas release by this stream will be negligible compared with the release via the main condenser exhaust. The turbine gland steam condenser exhaust samplers will be required by the Technical Specifications.
The staff has concluded that these provisions i
are acceptable.
.1 In the SER, the staft stated that certain items would be addressed when the Radiological Effluent Technical Specifications (RETS) were submitted by the i
applicant. These items were (1) sampling frequencies, required analyscs, instrument alarm / type setpoints, calibration, and sensitivities (2) frequency of routine instrument calibration, maintenance, and inspections (3) release rates The RETS have been submitted and, af ter extensive discussicn, have been revised' so that they are acceptable to both the applicant and the staff.
The RETS do address the aforementioned items and are considered acceptable by the staff.
In a May 20, 1986, letter to the staff, the applicant addressed the clarifica-tion of Items 4a and 4b of Attachment 1, " Noble Gas Effluent Monitor," to TMI Action Plan Item II.F.1, and clarification of Items 1 through 4 of Attachment 2,
" Sampling and Analysis of Effluents," to this same TMI Task Action Plan item.
On the basis of a review of this submittal and FSAR Sections 7.5, 11.5, and 12.3, the staff concludes that the applicant conforms with Attachments I and 2 except for the capability to obtain representative samples without excessive plateout. The applicant utilizes a system that has a very low flow velocity.
For those facilities that utilize this system, it has been incorporated as a l
license condition that, before the startup after the first refueling outage, the applicant must demonstrate that representative radioactive iodine and par-ticulata samples may be obtained.
This same license condition will be applied l
to the Seehrook Station.
The proposed license condition would read.as follows:
Before startup following the first refueling outage, the applicant shall demonstrate that the iodine / particulate sampling system is operable and will perform its intended function.
In Supplement No. 2, the staff identified certain items that would be reviewed as part of the review of the Offsite Dose Calculation Manual (ODCM).
The staff's review is addressed below.
Seabrook SSER S 11-3
In a letter to the applicant dated May 20, 1986, the staff agreed to the inclusion of certain tables in the Seabrook ODCM that are in the Technic Specifications of other plants.
These tables are in Part A of the Seabrook 00CM.
Part A cannot be changed without prior approval by the NRC.
idcntified by title and Standard Technical Specification (STS) number, are inThe tab Table 11.1 of this supplement.
l The. staff reviewed Part A of the ODCM submitted by letter (SBN-1122) dated l
June 17,1986.
The applicant had proposed numerous deviations from the Standard Technical Specifications.
After extensive discussions with applicant personnel, agreement was reached on mutually acceptable changes that meet both the intent of the Standard Technical Specifications and the plant-specific needs of Sea-brook.
The staff concludes that Part A of the Seabrook ODCM is acceptable.
Part B of the Seabrook ODCM, which constitutes the bulk of this ISO page manual, documents the methods used to meet the requirements of various Technical Speci-fications.
The methods addressed include the following:
(1) calculating offsite radiation doses from liquid effluents, gaseous effluents, and direct radiation i
(2)
\\
calculating offsite radiation dose rates from liquid and gaseous effluents (3) calculating setpoints for process, effluent, and engineered safety features instrumentation (4) performing the land use census (5) conducting the environmental monitoring program The staff has re ed the Seat, rook ODCM and noted the correction of the errors and deficiencies identified by the staff in the version submitted "for approval" by PSNH Letter / 54 (March 3,1986).
9 The staff also notes that the approaches used are generally consistent with NRC methodology and requirements.
not all differences between the -staff and the applicant have been resolvedHowever, either for Method 2 or for Method 1 gaseous releases.
The unresolved differ-ences in Method I are no greater than a factor of 10.
These differences are compensated by a requirement in Part A of the ODCM documented by a letter from the applicant dated July 22, 1986.
Part A requires that, pending resolution of the ODCM issues, only Method 1 (of Part B) will be used and that releases in gaseous effluents will be limited to no more than 10% of the quantities indi-cated to be permissible by Method 1.
ocrdh 20 and M 50 will not be exceeded.Thus, there is assurance that the limits s
ODCH acceptable % C r4 On this basis the staff finds the I
Seabrook SSER 5 11-4
)
1 (b) waste gas system (not required for accident centrol)
{
(6) Helium leak testing will be used for testing the gaseous systems.
(7) Action has been taken to reduce leakage potential from design and operator deficiencies as addressed in a generic letter dated October 17,1979.
The staff has reviewed this program description and concluded that it meets regulatory requirements and, therefore, is acceptable.
The applicant has r.'t yet submitted the results of leak rate measurements to tiie staf f.
This it required by NUREG-0737 before issuance of a full power license.
The applicant will be required to submit these data for staff review before 5% power is exceeded, and the license will be conditioned to state:
i Prior to proceeding above 5% of rated power, the applicant shall submit the results of leak rate measurements which demonstrate that its leakage reduction program has been successfully implemented.
By letter dated June 20, 1986, the applicant committed to submit these data to the staff before initial criticality.
1 Seabrook SSER 5 15-13
l
===.
Reason for Request===
Geometric configuration, permanent obstructions and/or structural interferences, prohibit 100% examination coverage of the Code-required volume on 2 welds and 100% examination coverage of the FSAR Augmented Examination volume on 7 welds listed in Table 1 of Relief Request PR-9.
Staf f Evaluation: The staff has reviewed the Applicant's submittal including Table I which identifies the welds for which relief is being requested, the Code Item Number, the examination angle and technique being used, the configuration of the weld, and the percent-age of the Code or FSAR Augmented Examination required volume that I
was examined.
Ir. addition, the Applicant reported that the subject welds received the ASME Code Section III volumetric examination during fabrication and the construction hydrostatic test.
Based on the above, the staff concludes that a significant percentage l
of the Code or FSAR Augmented Examination-required volumetric examina-tion has been performed and that these components would have to be redesigned in order to complete the remainder. Therefore, the staff concludes that the limited preservice volumetric examinations and the Section III fabrication examinations, along with the hydrostatic test, provide an acceptable level of preservice structural integrity and that compliance with the specific requirements of Section XI would result in hardship or unusual difficulties without a compensat-1 ing increase in the level of quality and safety.
1 IV.
CONCLUSIONS a
Based on the foregoing, pursuant to 10 CFR 50.55a(a)(3), the staff has detemined that certain Section XI required preservice examina-tions are impractical. The Applicant has demonstrated that either (i) the proposed alternatives would provide an acceptable level of quality and safety or (ii) compliance with the requirements would result in hardships or unusual difficulties without a compensating increase in the level of quality and safety.
The staff technical evaluation has not identified any practical method by which the existing Seabrook Nuclear Power Station, Unit 1 can meet all the specific preservice inspection requirements of Section XI of the ASME Code.
Compliance with all the exact Sec-j tion XI inspections would require the applicant to redesign a sig-nificant number of plant systems, obtain sufficient replacement components, install the new components, and repeat the preservice i
examination of these components.
Examples of components that would require redesign to meet the specific preservice examination provi-sions are:
the reactor pressure vessel, the reactor coolant pump I
discharge nozzle, and a number of the piping and component support systems. Even after the redesign efforts, complete compliance with the preservice examination requirements probably could not be achieved.
However, the as-built structural integrity of the existing primary pressure boundary has already been established by the construction code fabrication examinations.
Seabrook SSER 5 11 Appendix M i
I
A7.tachment 2 l
n: t closes 44 tT items. These items and the sections of this supplement that pre-9 sent results of the staff's evaluation are given below. Outstanding issues listed below that are discussed in sections followed by an asterisk become license conditions. Outstanding issues listed below that are discussed in sec-tions followed by a double asterisk become confirmatory items. '
(2) Emergency preparedness (2.3.3,** 13.3*)
\\
(4) Preservice and inservice inspection and testing programs (2.5.5.3, 3.9.6,**
5.2.4,* 6.6.3*)
(6) Environmental qualification of equipment (3.11**)
(8) TMI Action Plan items (4.4.5.4,"* 6.2.8, 9.3.4.2,13.3,* 14,* 15.9.9)
(10) Level measurement error (7.3.2.8)
(11) Instrumentation and control for safe shutdown (7.4.2.1,"* 7.4.2.4)
(12) Radiation data management system (7.5.2.2*)
(16) Shift Technical Advisor, TMI Action Plan Item I. A.1.1 (13.1.2.2,13.2.2.1)
As of this supplement, the remaining outstanding issues and the related SER sections are (6) Stress / dynamic qualification of equipment (3.10)
(14) Solid radwaste management system i
(20) Fire protection (9.5.1.9) 1.8 Confirmatory Issues Section 1.8 of the SER noted that there are some items that have been resolved essentially to the staff's satisfaction but for which certain confirmatory in-formation has not yet been provided by the applicant. This supplement closes 11 of the confirmatory itecs. These items and the sections of this supplement that present results of the staff's evaluation are given below. The confirma-tory issue listed below that is discussed in the section followed by an asterisk becomes a license condition.
The confirmatory issue listed below that is dis-cussed in the section followed by a double asterisk becomes an outstanding issue.
(1) Underground transmission line easement (2.1.2)
(12) Analysis of the containment purge and vent system (6.2.4, 6.2.8) l (13) Containment subcompartment analysis (6.2.1.2)
(14) Formal documentation of previously provided information related to several instrumentation and control systems (7.2.2, 7.4.2.5, 7.6.7.6)
Seabrook SSER 5 1-2
I i
Criterion (5) l l
The Seabrook service water system uses seawater for c:oling water, but the de-
.I 1
sign provides a double barrier between the primary co:lant system and the sea-water.
Grab sample analysis for chloride on diluted primary coolant liquid centrations in the liquid samples are determined by icn chgr le. Chloride con-y, which Vj has a detection limit of 10 parts per billion.
'-- ' r;,The sample can be diluted by as mucn as a factor of 15 and still provide the required minimum f
detectable threshold for chloride.
A N c^ E! --- " ht: ;;.a b, vy.,.uwwo en a a cuor vi mv iv.., b. o r n 2I
/p r:::;r.;il., ;.J.. ; ni'. (fif 7'J y.. rw,ca, -o kn wou a u mu i 6......L..w.
.:._u__ao g;
__ 7-
-m S[the PASS can also take an undiluted um.
sample 9hich would be retained for up to 30 days before analysis te pre +i4e for short-lived radioactivity decay, consistent with ALARA considerations.
e
}
The staff :oncludes that these provisions meet Criterion (5) in Item II.B.3 of NUREG-0737 and are, therefore, acceptable.
Criterion (6)
The applicant has performed a shielding analysis to er.sure that radiation ex-posure from reactor coolant and containment atmosphere sampling and analysis activities is within the acceptable limits of 5 rem wh:le body and 75 rem ex-tremities to the operator.
The operator exposure will include entering and leaving the sample panel area, operating the sample pa.el manual valves, per-forming manual sample dilutions, and transferring samples to the chemistry laboratory for analysis.
The staf f has determined that these provisions meet Criterion (6) in Item II.B.3 of NUREG-0737 and are, therefore, acceptable.
Criterion (10)
The accuracies, ranges, and sensitivities of the PASS instruments and analyt-i ical procedures are consistent with the recommendations of RG 1.97, Revision 3, and the clarifications of NUREG-0737, Item II.B.3.
Therefore, they are ade-quate for assessing the radiological and chemical status of. the reactor coolant.
The instrumentation was selected for its ability to operate in the postaccident sampling environment.
will be calibrated or tested at least once every 6 months. Equipment used Retraining of operators for postaccident sampling is scheduled at a frequency of once every 6 months.
The staff finds that these provisions meet Criterion (10) in Item II.B.3 of NUREG-0737 and are, therefore, acceptable.
Conclusion On the basis of the above evaluation, the staff concludes that the applicant's proposed PASS meets all 11 criteria in Item II.B.3 of NUREG-0737 and is, there-fore, acceptable.
The previously proposed license condition on the PASS :an now be removed.
Seabrook SSER 5 9-3
Dry active wastes will be processed in a 100-f t3 box compactor with antisprin back devices.
The compactor exhaust system operates to prevent dust from es g-caping during trash handling, and will be complete with fan, fan motor, pre-filter, absolute filter, and connections to the duct in the filled drum storage area exhaust system.
The applicant will have onsite storage capacity for 450 to 650 drums of solidi-fied waste.
Solid waste containers, shipping cask, and methods of packing are designed to meet applicable State and Federal regulations, including 10 CFR 71.
11.4.2 Evaluation and Findings The staff has reviewed the SWPS in accordance with the teceptance criteria of 1
SRP Section 11.4 (NUREG-0800). The scope of the review included line diagrams I
of the system, piping and instrumentation diagrams (P& ids), and the system description of the SWP5 and those auxiliary supporting systems that are essen-tial to the operation of the SVPS.
The applicant's proposed design criteria and design bases for the SWPS and the applicant's analysis of those criteria and bases have also been reviewed.
The staff has also reviewed (1) the capability of the proposed system to process the types and volumes of wastes expected during normal operation and anticipated operational occurrences in accordance with GDC 60 and (2) provision for the handling of wastes relative to the requirements of 10 CFR 20 and 71 and applicable U.S. Department of Trans-portation regulations.
>I The staff has reviewed the mobile solidification system that will be provided d*
by the NUS Corporation.
an NRC safety evaluation (Thomas, May 1985).This system was approved fo y
This evaluation stated that utili-zation of this system at a facility required that certain plant specific infor-i mation be supplied.
f},
The applicant did this in a letter dated May 29, 1986.
By letter dated June 17, 1986, spill control methods, including type and location.the applicant also provided a de The staff finds this acceptable.
s In addition, the mobile system waste processing will be limited to the waste processing building as noted in the May 29, 1986, submittal, and the waste i
classification requirements of 10 CFR 61 will be met.
The asphalt system process control program (PCP) has not been approved.
The [
PCP has not addressed j
(1) methods to verify that the waste has been solidified (2) methods for reprocessing the wastes when they are shown to be unsolidified J
The PCP must be approved before the processing of waste utilizing asp it.
~t The staff finds that both the mobile and inplant waste processinghms con-form with the remaining acceptance criteria of SRP Section 11.4.g....-..., :-
+'* '::':
.7 ^...:.,c. 6.,.:.. C.?' h..
- i i f :t ti; ;.c.f ta ;f R; Ib!$t' hI+ =
E'
, E,2
.3 2 f r '...,.. m.. '....
.. '. '. i._ w p roces s
^
h
..v i musu w >5e.
- inc a w..' t;; ::n
^
.u..
a.+t..
. a q,,. +., +
- g.,.7., <
- 7..... -. g g,,, g,,,,,,n; a
.aa....
st..
_,.u
- u. na.--a g ;7
.3 g n; w-Seabrook SSER 5 11-2
l l
I i'_' - :':"4+^ ^ 7 ^ 2 ^: # i., E:-..^',..m ouw> m eu vy cue en i aoo c.
4n 2 fn+nre e,.'
j
-enm a <
- tr
,,J,,,,
,,; a g;;+u
)
Tu fG n Q n + w :m 3r u id. In,"Sel.4' t -, !;f "ernfina 14enne^ 4e 4 e
d yn i
in the CFD b^fa--
r-
.,-c y Process and Effluent Radiological Monitoring and Sampling Systems s.,,+e g w
]
11.5
- t. e e m rad
{
1 11.5.2 Evaluation findings l
~
The staff questioned the lack of provision for sampling or monitoring the turbine gland steam condenser exhaust.
The applicant has committed to include i
(
provisions for continuously sampling this effluent stream for radiciodines and i
There is no noble gas monitor because the noble gas release by particulate.
this stream will be negligible compared with the release via the main condenser The turbine gland steam condenser exhaust samplers will be required exhaust.
by the Technical Specifications.
The staff has concluded that these provisions i
are acceptable.
In the SER, the staff stated that certain items would be addressed when the Radiological Effluent Technical Specifications (RETS) were submitted by the I
applicant. These items were (1) sampling f requencies, required analyses, instrument alarm / type setpoints, calibration, and sensitivities (2) frequency of routine instrument calibration, maintenance, and inspections l
l (3) release rates The RETS have been submitted and, after extensive discussion, have been revised' so that they are acceptable to both the applicant and the staff.
The RETS do address the aforementioned items and are considered acceptable by the staff.
In a May 20, 1986, letter to the staff, the applicant addressed the clarifica-tion of Items 4a and 4b of Attachment 1, " Noble Gas Effluent Monitor," to TMI Action Plan Item II.f.1, and clarification of Items 1 through 4 of Attachment 2,
" Sampling and Analysis of Effluents," to this same TMI Task Action Plan item.
i On the basis of a review of this submittal and FSAR Sections 7.5,11.5, and 12.3, the staff concludes that the applicant conforms with Attachments 1 and 2 except for the capability to obtain representative samples without excessive plateout. The applicant utilizes a system that has a very low flow velocity.
For those facilities that utilize this system, it has been incorporated as a license condition that, before the startup af ter the first refueling outage, the applicant must demonstrate that representative radioactive iodine and par-ticulate samples may be obtained.
This same license condition will be applied to the Seabrook Station.
The proposed license condition would read as follows:
Before startup following the first refueling outage, the applicant shall demonstrate that the iodine / particulate sampling system is operable and will perform its intended function.
In Supplement No. 2, the staff identified certain items that would be reviewed The as part of the review of the Offsite Dose Calculation Manual (ODCM).
staff's review is addressed below.
Seabrook SSER 5 11-3
,