ML19296D562
| ML19296D562 | |
| Person / Time | |
|---|---|
| Site: | Shoreham File:Long Island Lighting Company icon.png |
| Issue date: | 02/29/1980 |
| From: | Novarro J LONG ISLAND LIGHTING CO. |
| To: | Harold Denton Office of Nuclear Reactor Regulation |
| References | |
| SNRC-465, NUDOCS 8003050246 | |
| Download: ML19296D562 (41) | |
Text
~~
_ _ _ _ .l ^ i FLCO LONG ISLAND LIGHTING COM PANY
_ f. mg_aw SHOREHAM NUCLEAR POWER STATION P.O. DO X 618 NO RTH COU NTRY RO A D
- WA DING RIVL R, N.Y.11792 February 29, 1980 SNRC-465 Mr. Harold R. Denton, Director Office of Nuclear Reactor Regulation U. S. Nuclear Regulatory Commission Washington, D. C. 20555 Shoreham Nuclear Power Station - Unit 1 Doc.ket No. 50-322
Dear Mr. Denton:
Enclosed are fifteen (15) sets of responses to NRC information requests contained in letters sent to LILCO on March 19, June 14, and November 15, 1979. The specific responses transmitted with this letter are identified in the attached table of contents.
Responses to the remaining information requests contained in the referenced letters are being developed and will be forwarded to the NRC by the end of March, 1980.
i In addition, we have provided our positions concerning Shoreham SER Open Items 1 and 11. We trust that these responses will provide adequate resolution of your concerns.
Very truly yours,
- f. '
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0 cj P c v' A P g o J. . Novarro, Project Manager Shoreham Nuclear Power Station RAH /cc Enclosures
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SIIPS-1 FSAR RESI'0:!SE TO SECO:ID ROU::D REQUI'2TS E (SETS 22, 24 & 25)
Text, tables, and figures which are referenced in these responses, and attached, contain new or revised material. Text, tables, and figures referenced, but not attached, are existing r.aterial as presented in the BAR.
Docket flo. 50-322 February 29, 1980
S!iPS-1 FSAR TABLE OF CO?ITICITS Text. Tables, and Firmres 123-21 12.1-22 through 22c Figures 121.22-1 and 2 121-23 and 23a 121-24 121-25 122-5 3B-13 212-103 212-104 212-105 and 105a 331-24 333-25 through 25b 331-26 331-27 331-29 331-30 12./.-l and 2 Table 12.4.3-1 (2 pages) 420-39 through 39b 420-40 420-41 1
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REQUEST 121.21 We require that your inspection program for Class 1, 2 and 3 components be in accordance with the revised rules in 10 CFR Part 50, Section 50.55a, paragraph (q). Accordingly, submit the following information:
(1) A preservice inspection plan which is consistent with the required edition of the ASME Code. This inspection plan should include any exceptions you propose to the Code requirements.
(2) An inservice inspection plan submitted within six months of the anticipated date for commercial operation.
This preservice inspection plan will be required to support the safety evaluation eport finding regarding your comoliance with preservice and inservice inspection requirements. O' determination of your compliance will be based on:
(1) That edition of Section XI of the ASME Code referenced in your FSAR or later editions of Section XI referenced in the FEDERAL REGISTER that you may elect to apply.
(2) All augmented examinations established by the Commission when added assurance of structural reliability was deemed necessary. Examples or augmented examination requirements can be found in the NRC positions on: (1) hich energy fluid systems in Section 3.6 of the Standard Review Plan (SRP), NUREG-75/087; (2) turbine disk integrity in Section 10.2.3 of the oRP; (3) the BWR feedwater inlet nozzle inner radii, UUREG-312, and (4) BWR SS piping NUREG-313.
Your response to this item should define the applicable edition (s) and subsections of Section XI of the ASME Code.
If any of the examination requirements of the particular edition of Section XI you referenced in the FSAR c(nnot be met, a request for relief must be submitted, including complete technical justification to support your request.
Detailed guidelines for the preparation and content of the inspection programs to be submitted for staff review and for relief requests are attached as Appendix A to Section 121.0 of our review questions.
RESPONSE
A Preservice Inspection Plan was submitted via Lilco letter SNRC-265 to Mr. Karl Kniel dated March 16, 1978. An Inservice Inspection Plan will be sub.nitted six months prior to the date of commercial operation.
121-21
REQUEST 121.22 Supply the pressure-temperature limits for SNPS-1 for the following conditions:
(1) Heatup-Cooldown (2) Normal Operations (3) Inservice Leak and Hydrostatic Tests Specify the number of effective full power years that these limits are to be applicable, the predicted fluence of the quarter thickness wall locatica, the radius and thickness of the reactor vessel, and any assumptions used in calculating those limits.
RESPONSE
Operating limits which define minimum reactor vessel metal temperature vs. reactor pressure during normal operation heatup-cooldown and inservice leak and hydrostatic testing are based on fracture toughness and were established using the methods of Appendix G of Section III of the ASME Boiler and Pressure Vessel Code, 1971 Edition (Su::rer 197 2 Addenda ) .
These limits are shown on Figure 121.22-1. The limit curves presented in this figure are applicable for the life of the vessel. The predicted flugnce of the quarter thickness wall location is 4.4 x 1018 n/cm . The reactor vessel radius is 110 inches with a minimum wall thickness of 5.38 inches at the core beltline.
All the vessel shell and head areas remote from discontinuities, plus the feedwater nozzles, were evaluated. The operating limit curves are based on the limiting location. The boltup limits for the flange and adjacent shell region are based on a minimum metal temperature of RTNDT + 60 F. The maximum through-wall temperature gradient from continuous heating or cooling at 100 F per hour was considered. The safety factors applied were as specified in ASME Code Appendix G and GE Licensing Topical Report NEDO-21778-A.
Operating Limits Durinc Heatup, Cooldown and Normal Operation The fracture toughness analysis was performed for the normal heatup or cooldown rate of 100 F/ hour. The temperature gradients and thermal stress effects corresponding to this rate were included.
The results of the analyses are a set of operating limits for 121-22a
non-nuclear heatup or cooldown shown as curves labeled B and B' on Figure 121.22-1. Curves labeled C and C' or. these figures apply whenever the core is critical. The basis for curves C and C' is described in GE BWR Licensing Topical Report NEDO-21778-A.
Temperature Limits for Preoperational System Hydrostatic Tests and ISI Hydrostatic or Leak Pressure Tests Based on 10CFR50 Appendix G IV.A.2.d, which allows a reduced safety factor for tests prior to fuel loading, the preoperational system hydrostatic test at 1250 psig may be performed at a minimum tenperature of 91 F which is established by the RT N DT of the bottom head plate plus 60 F. The fracture toughness analysis for system pressure tests resulted in the curves labeled A and A' shown on Figure 121.22-1. The curves labeled
" Core beltline" are bac 2d on - initial RTNDT of 19 F.
The beltline weld mate ial is expected to be more limiting at end-of-service fluence levels, and this weld material has an initial RTNDT of -50 F. The predicted shift in the RTNDT from Picure 121.22-2 (based on the neutron fluence at 1/4 of the vessel we.il thickness) must be added to the beltline curve to accoant for the cffect of fast neutrons.
Temperature Limits for Boltup A minimum temperature of 70 F is required for the closure studs.
A sufficient number of studs may be tensioned at 70 F to seal the closure flange 0-rings for the purpose of raising reactor water level above the closure flanges in order to accist in warning them. The flanges and adjacent shell are required to be warmed to minimum temperatures of 70 F before they are stressed by the full intended bolt preload. The fully pre-loaded boltup lindts are shown on Figure 121.22-1.
Operating Limits Based on Fracture Toughness For the purpose of setting these operating limits, the reference temperature, RTNDT, is determined from the toughness test data taken in accordance with requirements of the Code to which this vessel is designed and manufactured. These toughness test data,
. Charpy V-notch (CVN) and/or dropweight nil ductility transition temperature (NDT) were analyzed to permit compliance with the intent of 10 CFR 50 Appendix G. Because all toughness testing needed for strict compliance with Appendix G was not required at the time of vessel procurement, some toughness results are not available.
121-22b
For example, longitudinal charpy V notch (CVN) instead of transverse, were tested, sometimes at only a single test temperature of +10 F for absorbed energy. Also, at the time either CVU or NDT testing was permitted; therefore, in many cases both tests were not performed as is currently required.
To substitute for this absence of certain data, toughness property correlations were derived for the vessel materials in order to operate upon the available data to give a conservative estimate of RTNDT, compliant with the intent of 10 CFR 50 Appendix G criteria.
These toughness correlations vary, depending upon the specific material analyzed, and were derived frem the results of NRC Bulletin 217, " Properties of iteavy Section Nuclear Reactor Steels", and from toughness data from the Shoreham vessel and other reactors. In the case of vessel plate material (SA-533 Grade B, Class 1), the predicted limiting toughness property is either NDT or transverse CVN transition curve results and NDT values are available for all Shoreham vessel plates. The transverse CVN 50 ft-lb transition temperature is estimated from longitudinal CVN data in the following manner. The lowest longitudinal CVN ft-lb value is adjusted to derive a longitudinal CVN 50 ft-lb transition temperature by adding 2 P per ft-lb to the test temperature. If the actual data equal or exceed 50 ft-lb, the test temperature is used. If sufficient data are available as in the case of Shoreham, the 50 ft-lb tempcrature is derived by interpolation. Once the long. odinal 50 ft-lb temperature is derived, an additional 30 F added to account for orientation effects and to estimace the transverse CVN 50 ft-lb temperature minus 60 F, estimated in the preceding manner.
For forgings (SA-508 Class 2), the predicted limiting property is the same as for the vessel plates. Both NDT and CVN values are available for the vessel flange, closure head flange, and feedwater nozzle materials for Shoreham. These forging RTNDT values are derived from the data in the same way as for the vessel plates.
For the vessel weld metal, the predicted limiting property is the CVN 50 ft-lb transition temperature minus 60 F, as the NDT values are -50 F or lower for these materials. This temperature is derived in the same way as for the vessel plate material, except the 30 F addition for orientation effects is omitted since there is no principal working direction. When NDT values are available, they are also considered and the RTNDT is taken as the higher of NDT or the 50 ft-lb temperature minus 60 F.
When NDT is not available, the RTNDT shall not be less than
-50 F since lower values are not supported by the correlation data.
121-22c
For vessel weld heat affected zone (HAZ) material, the RT NDT is assumed the came as for the base material since ASME Code weld procedure qualification test requirements and post weld heat treatment indicate this assumption is valid.
Closure bolting material toughness test requirements for Shoreham were for 30 ft-lb at 60 F below the bolt-up temperature.
Current Code requirements are for 45 fr-lb and 25 mils lateral expansion (MLE) at the preload or lowest service temperature, including bolt-up. Therefore, since CVN values as low as 40 ft-lb exist for Shoreham closure bolts, 60 F is added to the test temperature in order to derive the bolt-up temperature.
Using the above general approach, an initial RTUDT of 19 F was established for the core beltline region for Shoreham Unit No. 1.
The effect of the main closure flange discontinuity was considered by adding 60 F to the RTNDT to establish the minimum temperature for boltup and pressurization. The minimum boltup temperature of 70 F for Shoreham Unit 1 is based on an initial RTNDT of + 10 F for the closure flange forgings.
The effect of the feedwater nozzle discontinuities were considered by adjusting the results of a BWR/6 reactor discontinuity analysis to the Shoreham reactor. The adjustment was made by increasing the minimum temperatures required by the difference between the Shoreham and BWR/6 feedwater nozzle forging RTNDT's. The feedwater nozzle adjustment was based on an RTNDT of -20 F.
The reactor vessel closure studs have a minimum Charpy impact energy of 40 f t-lbs and a 23 mil lateral e::pansion at 10 F for Shoreham. The lowest service temperature for boltup is taken to be 60 F above the 10 F values or 70 F.
121-22d
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6 REQUEST 121.23 To demonstrate compliance with Appendix H to 10CFR Part 50, include in the SNPS-1 FSAR and Technical Specifications a table that provides the following information for each surveillance specimen capsule:
(1) The actual surveillance materials in each capsule.
(2) The beltline material from which each surveillance material was obtained.
(3) The test specimen type (s), and their orientation, for each surveillance material.
(4) The actual location of each capsule in the reactor vessel.
(5) The lead factor for each capsule calculated with respect to the 1/4 wall thickness.
(6) The proposed loading schedule of the capsules into the SNPS-1 reactor vessel.
(7) The proposed time of capsule withdrawal (calecdar years and effective full power years).
Also, state the ASME Code requirement (specify edition, addenda, section and paragraph), the ASTM Standard (specify number and year),
or any other document from which the materials surveillance program for SNPS-1 was formulated.
RESPONSE
(1) There are Charpy V-Notch and tensile test specimens from base metal, weld metal, and weld heat affected zone (HAZ) contained in each of the 3 capsulas. See FSAR Section 5.2.4.4. These specimens are as described in parts (2) and (3) below.
(2) The surveillance specimen test weldment was fabricated by welding two plates from Heat No. C4882. These two plates are identified as C4882-1 and C4882-2; however, since.they are the same heat, they have similar chemical composition and mechanical properties. Weld filler metal was from Type B4, Heat 1P3571, with flux type Linde 1092, lot 3958; and type E-8018C3, lots EODJ and KACI. Longitudinal base metal Charpy specimens for baseline and surveillance were taken from C4882-1. HAZ Charpy and tensile specimens were also taken from the C4882-1 side of the weld. Base metal tensile specimens and transverse baseline C arpy specimens were taken from C4882-2. Quality Assurance records for the heat numbers which identify the materials that make up the steel are kept by General Electric and on the site.
121-23a
(3) The type and orientation for each surveillance material are as follows:
Type Orientation Wold Metal (weld direction Tensile specimens parallel ta perpendicular to base plate weld direction, in weld principal rolling directions)
Charpy specimens perpendicular to weld direction, notch centered in weld HAZ Tensile and Charpy specimens perpendicular to weld direction and parallel to base plate principal rolling direction, centered on HAZ of C4882-1 plate near fusion line.
Base Metal Tensile and Charpy specimens parallel to principal rolling direction.
(4) The capsules are positioned at core midline and at the azimuthal locations of 30, 120, and 300.
(5) The factor to obtain 1/4 wall thickness location is 1.26. This factor includes a decrease to adjust for the radial displacement from the centerline of the capsule to the 1/4 wall thickness location and an increase to adjust for the variation in flux from the 30 degree azimuthal location to the peak power location.
The caprules are radially equidistant from the core centerline and sym.netrically spaced with respect to the peak power location.
Therefore, the lead factor for each capsule is the same.
(6) The capsule containing the specimens will be loaded into the reactor pressure vessel prior to fuel load.
(7) The proposed time of capsule withdrawal is to be in accordance with 10CPR50 Appendix H paragraph II.C.3.a, utilizing the 3 capsules described in 1 above. See FSAR Section 5.2.4.4.3.
This is proposed even though the predicted end-of-life toughness transition temperature in accordance with Section IIIB is greater than 100 F but less than 200 F, since the program was designed prior to 10CFR50 Appendix H.
The SNPS-1 Surveillance Program is based on the ASTM E185-70 standard and Section III of the ASME code 1965 edition through Summer 1966 addenda as applicable to the materials used. Section 5.2.4 of the FSAR will be revised to incorporate the information presented above.
The number of surveillance capsules and the frequency for withdrawal will be included in the Technical Specifications.
121-23b
REQUEST 121.24 It is our position that the construction and inspection of the reactor surveillance capsule attachments be done according to the requirements for permanent structural attachments to the reactor vessels specified in ASME Code Sections III and XI.
RESPONSE
The hermetically sealed capsules are not attached to the vessel but are in welded capsule holders. The capsule holders are mechanically retained by capsule holder brackets welded to the vessel cladding. The capsule holder brackets allow the capsule holder to be removed at any desired time in the life of the plant for specimen testing. These brackets are designed, fabricated, and analyzed to the requirements of ASME Code Section III.
The capsule holder brackets will be inspected in accordance with ASME Section XI as noted in the Shoreham Preservice Inspection Program Plan.
The Shoreham Nuclear Power Station Inservice Inspection Program is currently being developed based on the 1977 Edition through Summer '78 Addenda of the ASME B&PV code with modifications on November in conformance with 10CFR 50.55a as amended 1, 1979. In conformance with ASME Section XI, Table IWB-2500-1, Examination Categories B-N-1 through B-N-3, accessible welds of interior attachments to the reactor pressure vessel will be visually examined during each inspection interval. This requirement will be applied to the accessible portion of the bracket welds to the vessel cladding.
121-24
REQUEST 121.25 Specify the fracture toughness requirements imposed on the ferritic materials used in the piping, pumps and valves, and bolting over one' inch nominal diameter, used in the reactor coolant pressure boundary of SNPS-1.
RESPONSE
There are no ferritic pumps in the SNPS-1 reactor coolant pressure boundary. The fracture toughness requirements for the valves, piping and bolting over one inch nominal diameter used in the S:;PS-1 reactor coolant pressure boundary are as follows:
NSSS SCOPE MSIVs Fracture toughness requirements in keeping with the 1968 ASME Nuclear Pump and Valve Code, Winter 1969 Addenda draft.
SRVs Fracture toughness requirements in keeping with the 1968 ASME Nuc? car Pump and Valve Code, Summer 1970 Addenda.
Main Steam Main steamline piping was procured to the requirements Piping of 1967 ANSI B31.1 which did not require toughness testing.
Bolting Closure studs were tested at the preload, or the lowest service temperature (70F minus 60*F to give lowest values of 40 f t-lb nini:'m Charpy V-notch impact energy and 23 mils lateral expansion at +10 F.
BOP SCOPE ASME III Section NB-2300 of the ASME III code, 1971 edition, Class 1 winter 1972 addenda.
Piping and Valves 121-25
SNPS-1 FSAR Request 122.5:
We have determined that revision 1 (July , 1978) to Regulatory Guide 1.56, aMaintenance of Water Purity in boiling Water Reactors," is applicable to the Shoreham Station. Modity your FSAR to describe how the revised Regulatory Guide will be implemented.
Response
The plant design complies with Revision 1 (July 1978) to Regulatory Guide 1.56 " Maintenance of Water Purity in Bolling Water Reactors."
f 122-5
,f
SNPS-1 FSAR 3B-1.52 Desian , Testina, and Maintenance Criteria for Atmonohere Cleanup, nyatom Air Filt res *.lon , and Isnnorotion Units or Light-Water-Cooled Nuclear Power Pignts (6/ 7 3)
The Reactor building Standby Ventilation System 111ter trains and the Control Room Air Conditioning Illter trains are in compliance with Regulatory Guide 1.52, except that they are not removatie as single units.
Reference Section 6.2.3 38-1.53 Aoolication of the Sincle-Failure Criterion to Nuclear Power Plant Protection Systems (6/73)
The application of the single failure criterion to the protection system complies with Regulatory Guide 1.53.
Reterence Section 7.3.2 3B-1.54 Quality assurance Recuiren. ants for Protective Coatinas Applied to Water-Cooled Nuclear Power Plants (6/73)
Quality assurance requirements for protective coatings comply with Regulatory Guide 1.54.
e Reference Sections 6.2.1 and 6.3.2 and Chapter 17 The Operational QA Program conforms to the guidance provided in Regulatory Guide 1.54 dated 6/73 during the operational phase.
3B-1.55 Concrete Placement in Cateoory I Structures (6/73)
Concrete placement in Category I structures complies witn Regulatory Guide 1.55 except that all reinforcing shop detail drawings other than for the containment mat, pedestal, primary wall and drywell wall are checked by engineers at the jobsite.
Reference Section 3.8 and Chapter 17 3B-1.56 Maintenance or Water Purity in Boiling Water Reactors (7/78) l The plant design complies with Regulatory Guide 1.56. Operating limits are contained in the technical specifications.
Reference Sections 5.2.3.4 and 10.4.6 and Chapter 16 33-1.57 Design Limits and loadina Combinations for Metal Primary Reactor Contaltunent System Components (c/73)
Regulatory Guide 1.57 is not applicable to this plant, since the primary containment design is not the metal primary reactor containment type.
SNPS-1 FSAR Request 212.103 (PSP) (15.0) :
The analyses or abnormal operational transients in tne FSAR utilize the REDY computer code described in General Electric Company Topical Report NEDO-10802, " Analytical Evaluations for the General Electric Boiling Water heactor." Recent conrirmatory tests conducted at Peach Bottom Unit 2, however, revealed tnat in certain cases, the results predicted by REDY are nonconservative.
Therefore, we require the applicant to recalculate the minimum critical power ratio for the limiting transient (generator load rejection without bypass) using General clectric Company's new computer code ODYN. We will use tne results of tnis calculation to verify the acceptability of the minimum critical power ratio.
Response
Although an NRC letter has been issued approving the ODYN model, generic discussions between GE and the NRC are continuing relative to code input methods and the need, it any, to apply an additional conservatism f actor in establishing the MCPR operating limit. An ODYN reanalysis of the limiting transient (generator load rejection witnout bypass) will De conducted after tnese generic issues have been resolved between GE and the NRC and the results will be submitted accordingly. .
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SNPS-1 FSAR Request 212.104 (RSP) (15. 0) :
In analyzing anticipated operational transients, the applicant has taken credit for plant operating equipment which has not been shown to be reliable as required by General Design Criterlon 29.
The starr has discussed the application ot this equipment generically with General Electric. In these discussions General Electric has stated that the most limiting transient that takes credit for this equipment is the excess 1eedwater event.
Further, General Electric has stated that the only plant operating equipment that plays a signiticant role in mitigating this event is the turbine bypass system and tne Level 6 nigh water level trip (closes turbine stop valves) . We will allow the use ot the turbine bypass and Level 8 high watcr level trip systems in mitigating transients except for the turbine trap and generator load rejection without bypass transients which are currently minimien critical power ratio-limiting.
To assure an acceptable level of perrormance, it is the staff's position that this equipment be identiried in the plant Technical Spec 1tications with regard to availability, set points, and surveillance testing. The applicant must submit his plan for implementing this requirement along with any system moditications that may be requirea to fulfill this requirement.
Response
In discussions between GE and the NRC on Novenber 20 and 21, 1978, GE presented the results of transient analysis performed to design 1 sis accident condition assumptions (i .e . ,
equipment availability) The analysis indicated that railure to give credit to tne mvel 8 turbine trip and the main turbine bypass system could result in a ditierence in the critical power ratios ot 0.02 and 0.08, respectively. Therefore, these postulated conditions could not result in unacceptable impacts on the health and satety or the public.
The Level 8 instrumentation will be sub 3 ect to technical speci11 cation requirements associated with the HPCI system. The proposed technical specifications for the racility will address this concern.
The turbine steam bypass system and stop valves are turnished with the main turbine generator and have exhibited high reliability on existing nuclear and fossil rueled operating units.
Normal operating procedures require that the valves be functionally exelcised periodically in accordance with vendor recommendations. This will ensure valve operability and provide adequate assurance that the valves will operate when required.
212-104 s .u7 -
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SNPS-1 FSAR Roonest 212.105 (RSP) (15.0) :
We consider ATHS to be an unresolved safety issue. However, we have described the type of plant moditications which, it provided, would reduce ATWS risk to an acceptable level.
Volume 3 of NUREG-04t>0 which describes the rationale for specitying these plant modir2 cations is being reviewed by the Advisory Committee on Reactor Sateguards. The hegulatory Requirements Review Connittee has completed its review and concurred with our approach described in Volume 3 of NORt.G-0460 insolar as it applies to Shoreham. We plan to issue requests for the industry to supply generic analyses or ATWS mitigation capability and anticipate presenting to the Commission in May 1979 our reconnendations for its actions to resolve the NrWS concern. Shoreham would be required to implement plant modirications in conformance with the Commission's rinal resolution on this issue.
We require that the applicant agree to implement modifications on a schedular basis in coniormance with the Concnission 's rinal resolution of this issue.
In the event that Shoreham starts operation berore necessary plant modifications are implemented, we require some Interim actions be tanen by the applicant in order to reduce, turther, the risk rrom ATWS events. Tne applicant is required to:
(1) Shoreham must have an acceptable recirculation pump trip to assure that the short-term consequences of ATUS evento do not result in excessive primary system overpressurization. The criteria Ior an acceptable recirculation pump trip design are specified in Appendix C of Volume 3 of NUREG-04o0 The recirculation pump trip designs for Zimmer, Monticello, and Hatch (modified) have Deen found acceptable.
(2) Develop emergency procedures to train operators to recognize an ATWS event, including consideration of scram indicators, rod position indicators, flux monitors, vessel level and pressure Indicators, rellet valve and isolation valve indicators, and containment temperature, . pressure, and radiation indicators.
(3) Train operators to take actions in the event or an ATWS including consideration of immediate manual scrarmning or the reactor by using the manual scram buttoms rollowed by changing rod scram switches to the scram position, stripping the feeder b. a};ers on the reactor protection system power dis tributi n buses, opening the scram discharge volume drain valve, prompt actuation of the standby liquid control system, and prompt placement of the residual heat removal system in the pool cooling mode to reduce the severity or the containment conditions.
212-105
SNPS-1 FSAR Early operator action as described above, in conjunction with a recirculation pump trip, would provide signiticant protection for some ATWS events, namely those which occur (1) as a result or co:amon n: ode t allure in the electrical portion of the scram systeu and some portions 01 the drave system, anu (2) at low pcuer levels where the existing standby liquid control system capahility is sutficient to limit the pool temperature rise to an acceptable level.
Response
The following interim actions will be amplemented at Shoreha:a in order to rurther reduce the risk associated with anticipated 5
tranutents witn failure to scram (ATWS) events:
- 1. A recctor recirculation pump trip system will be imple:nented at Shoreha:a unich meets the criteria f or an acceptacle recirculation pump trip cesign as specirled in Appendix C of Volurae 3 ot NURLG-0460, "Anticipaten Transients Without Scram for Light Water Reactors."
- 2. Emergency procecures will be devclooed tor ATWS events.
These procedures will be similar to energency proceaurus developed tor use at Shoreham whicn consist or tne following six sections:
(a) Symptoms (b) Automatic actions (c) Immediate actions (d) Subsequent actions (e) Final conditions (f) Discussion Operators will be trained to parform the proper actions Ior ATWS events as part of the f ormal operator training progra:a.
It is anticipated these procedures will be completed and submitted approximately 6 months prior to the tuel load date.
This was previously addressed in SNRC-437, dated October 19, 1979.
212-105a
SNPS-1 FSAR Request 331.24 ( 12.3) :
Describe permanent shielding provided to assure acceptanle radiation levels in potentially occupied areas in the vicinity or the spent tuel transfer process. If very high radiation areas are projected, describe precautions taken to prevent inadvertent personnel access during fuel transrer.
Response
Permanent shielding, consisting or a6ft thick concrete wall structure surrounding the reactor vessel snroud and tne s pent fuel pool, assures acceptable radiation levels in the reactor building during fuel transfer (see Fig. 12.3.1-16) . ,
In the drywell, the 2 ft thick concrete sacriticial shield assures that acceptable radiation levels exist in th e general area surrounding the sacrificial shield. In the specific area of the reactor pressure vessel nozzle penetrations, administrative controls will be used to assure that no personnel will nave access in the immediate area of the reactor pressure vessel ;
nozzles during the removal of the fuel asse: ably trom the reactor ;
pressure vessel. <
l A portable 6 in. lead bottom rerueling radiation shield, whien dL spans the inner and outer rerueling Lellows area between the rerueling canal and the reactor pressure vessel riange curing spent tuel assembly transit, will be used to reduce radiation levels in the drywell.
Y 1
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l 1
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1 331-24 ;
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SNPS-1 FSAR Request 331.25 ( 12.0) :
Describe how the tollowing Regulatory Guides have neen followed at Shoreham and, it not rollowed, describe tne spec 1ric alternative methods used:
R.G. 8.2, " Guide for Administrative Practices in Radiation Monitoring" R.G. 8.3, " Film Badge Performance Criteria" R.G. 8.4, " Direct-Reading and Indirect-Reading Pocket Dosimeters" R.G. 8.7, " Occupational Radiation Exposure Records System" R.G. 8.9, " Acceptable Concepts, Models, Equations, and Assumptions f or a Bioassay Program" R.G. 8.12, " Criticality Acci'- 9t Alarm Systems" ;.
R.G. 8.14, " Personnel Neutrol Dosimeters" j R.G. 8.15, " Acceptable Programs for Respiratory Protection" ]
g R.G. 8.19, " Occupational Radiation Dose Assessment in Light Water Reactor Power Plants Design Stage Man-Rem Estimates" Resoonse:
The radiation monitoring program at Shoreham is structured within I the guidelines or this Regulatory Guide such tnat resultant ;
radiation exposures and releases of radioactive materials in ;
effluents to unrestricted are as are maintained as low as i reasonably achievanle. We have followed the concept or obtaining 1 the services of a qualified expert as required curing tne early planning and engi neering phases. Funagement will continue to l evaluate this need during the operation phase. Records will be ;
maintained in accordance with the requirements or 10CFR20.401. ;
Film badges will not be used for personnel monitoring at l Shoreham. Thererore, no need exists to follow this Regulatory ,
Guide. .
Regulatory Guide 8.4 Specirications cited in ANSI N13.5-1972 will be met or exceeded by the Direct-Reading Pocket Dos 1 meters ultimately selected tor j use at Shoreham. Station Health Physics procedures address the a
/
331-25 l i
i
SNPS-1 FSAR Regulatory Guide position on types or training, trequency of testing, and standards of rejection. Since Direct-Reading Dosimeters will not be used at Shoreham for the determination or neutron-to gamma ratios, this portion or tne Regulatory Gulce is not addressed.
Regulatory Guide 8.7 The plant's Dose Records System is presently being organized utilizing the applicable guides and standards. Station Health Physics procecures and/or computer sottware address the subjects of record keeping for training, external personnel dose, whole-body counter records, unusual exposure, damaged or lost dosimetry records, instrument calibration and maintenance, and radiation status or work areas. Since medical radiation exposures are not part of occupational radiation exposure, no records are maintained. Retention and storage or records will meet the applicable requirements.
Regulatory Guide 8.9 This Regulatory Guide is insufficient for present cay Bloassay programs. Shoreham is structuring its whole body counting program with tne guidance of ANSI N343 on Internal Dosimetry.
Measurements of radioactive material in excreta will not be done on a normal basis.
Regulatorv Guide 8.12 This Regulatory Guide addresses the requirements or 10CFR70.24 regarding criticality accident alarm systems. Our new fuel license application nas requested an exemption from Section 70.24 for hanaling anc storage of new ruel elements. A secono detector will be added in those areas where accident criticality monitoring is necessary if the exemption is denied.
Regulatory Guide 8.14 Shoreham plans to use TLDs for neutron dosimetry. These TLDs shall be selected so that they meet the intent or ANSI N319-1976 and Regulatory Guide 6.14. Until such time as TLD neutron dosimetry is obtained, Shoreham will calculate neutron dose equivalent in place or neutron dosimetry using a neutron-to gamma ratio obtained by the use of a portable survey instrument.
Regulatory Guide 8.15 Station Health Physics procedures address the Respiratory Protection Program at Shoreham. They have been tailored in accordance with Regulatory Guide 8.15 and NuknG-0041. In addition, the Company's general otfice Medical Depa rtment is meeting the requirements for a Medical Program. A written policy statement or respirator usage will be added to the FSAR.
331-25a
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I
SNPS-1 FSAR Regulatory Guido 8.19 This Regulatory Guide deals with design stages and edrly construction phases of the plant. Our position is stated in the response to Request 331.30.
c 331-25b
StiPS-1 FSAR Request 331.26 (12.1.2 ) :
Describe your use of wall and floor coatings in cubicles containing radioactive equipment to f acilitate decontantination in the event of a spill.
Response
In all potentially contaminated areas, all floors and wall surfaces (ror a minimum height of 8 feet) will have protect 1ve coatings applied. The coating systems will provide a uniform sealed surtace for all coated substrates. The sealed surraces will racilitate decontainination or the area in the event or a spill.
c 331-26
SMPS-1 FGAR Request 331.27 (12.3.4):
In erder to alequately detect alrhorne radioactivity in areas which may be occupied by personnel, airborne radioactivity monitors should be located upstream or the air cleaning systems.
It does not appear from Figure 9.4.3-1A that you have provided a radiation detector upstream ot the air cleaning system tor the radw a ste building. Provide assurance that all airborne radioactivity monitors sampling air from areas whicn may be occupied by personnel have sampling points upstream or tne air cleaning systems. Also justify why you have removed the radiation detector trom the line for the reactor building standby vent system in Figure 9.4.2-1, Rev. 16.
Response
Tne main function of the detector / sample system shown on Figure 9.4.3-1A is to aid the operator in ascertaining the radioactivity con tribution from the radwaste building to the total erfluent release rate Ircm the Station Ventilation Exhaust, with a seconda ry function of indicating any gross enange in airoorne activity (Indicatea directly tor noble gases and indirectly for particulates by deviation f rom an establisned normal level during plant operation) . The detector location is selected for best performance of its primary function.
The provisions ror directly measuring airborne levels in a specitic machinery cubicle atmospnere or general access area which may be occupied by personnel will be via portaole continuous alrDorne activity monitoring system. These portable systems will have collection and readout capability. This is described in more detail 'n FSAR . Sections 12.3.4.2 and 12.5.2.2.2. c The radiation detector has not been removed from the line for the reactor nullding standby ventilation system (RBdVS). When Figure 9.4.2-1 (reactor building nonnal ventilation system) was revised it did not contain all the details of the RBSVS. The RBSVS is shown in more detall on Figure 6.2.3-1 and the radiation detector is still on the line.
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331-27
SNPS-1 FSAR Requent 331.29 ( 12. 3) :
Justify your placement of the rollowing multiple components within a single cubicle as meeting the ALARA pnllosophy: 1)
Cation regeneration tank, Anion regeneration tank, and regenerative resin storage tank in a single zone VI cunicle on the 15'-0" level or the radwaste building: and 2) Eight pumps in a single zone IV cuolcle on tne 15'-0" level or the radwaste building. Describe any plans to use portaDle shielding in areas not provided with permanent shield walls.
Response
The placement or the Cation Regeneration Tank, Anion Regeneration Tank, and Regenerative Resin Storage TanA in a single cunicle is not inconsistent with ALARA philosopny. Tanks and vessels are passive components not requiring extensive normal maintenance and tanks of the same suDsystem are normally not shielced Irom each other. If it is necessary to perfona maintenance on one or the tanks, all three will be drained and flusned, it necessary, to reduce dose rates to acceptable levels within tne cuolcle.
There are six pumps located in a single cubicle on el 15-6 of the radwaste building. Also, there is a space reservation for two i additional pumps in that cubicle. The six pumps are listed below:
- 1. Regen Evap Feed Pump No. 1
- 2. Regen Evap Feed Pump No. 2
- 3. Floor Drain Collector Tank Pump No. 1 '
- 4. Floor Drain Collector Tank Pump No. 2 !
- 5. Waste Collector Tank Pump No. 1 j
- b. Waste Collector Tank Pump No~. 2 Each of these pumps is separated from t2. . radioactive tann associated with it in accordance with the ALARA principle or l separating active components requiring maintenance trom large ;
radioactive sources. These are small pumps that are not expected j to experience continuous duty. Only minor maintenance will be performed on the pumps in place. For extensive repairs, a pump will be removed to a clean area thereby limiting exposure to e personnel. I In accordance with the ALARA philosophy or Regulatory Guide 8.8 portable shielding may be used in areas not provided witn a permanent shield walls. Berore entering radiation areas where signiricant doses could be received, station nealth physics personnel will monitor radiation levels. The decision to provide :
portable shielding will be based on such. tactors as the estimated '
exposure time required to complete a task and the estimated doses anticipated trom the exposure versus the time required to install portable shielding. In conclusion, the equipment arrangement described is not inconsistent with the ALARA philosophy.
331-29
RI,0UEST 331.30 Using recent data from operating plants similar in design to Shoreham, provide a more detailed bicakdown on the estimated annual doses by Job category at Sho re harr . These estimates should include: 1. the tanks to be performed during operation and anticipated occurrences: 2. the time and manpower required to perform these tasks; and 3. the expected dose rates to which workers will be exposed in performing those tasks. Regulatory Guide 8.19 prov. ides guidance in making such an assessment.
_R_E S P O N S U The response is incorporated in revised I'SAR Section 12.4.3 and Tabic 12.4.3-1 attached.
331-30
StiPS-1 FSAR 12.4 DOSE ASSESSMENT 12.4.1 Dn.icn ob9ctiven The design or the rhielding is based on conservative estimates of occupancy time required .tn each area of the plant. An errort has inen made to keep the dose to plant personnel as low as practicable. This is accomplished by an extrenely conservative approach to the design or chielding for each area. Since adequate conservatin, is erployed in the cource and chield modeling, no acditional conservatisn is introduced in setting radjation zone recuirenants. Table 12.4.1-1 lists the six sono designations which have Men entablished alonc with the maximum allowable doce rate and estimated occunancy tin.a for each area.
Due to variations in onorational r' odes and unknowns associated with equipment Ic.aintenance requirements , it is not possible to detinitely specily occupancy times in any given area of the plant.
12.4.2 Airborne Acti"ity Due to conservative estimates of equipment leakage rates, it is postulated that certain ureus of the plant will have some airborne a ct iv i t',, concentrations as discussca in detail in Section 12.2.2. The airborne activity concentration in the turbine building resulti:n tron design basis failed fuel levels and leak rates are calculated to be 25 percent or the :raxinum per mi u.ibl e ecncentrations (DC) for occuc tional workers as detined in 10CFE20, Apptadix 13, colum 1. For the reactor builling, the concentrat. ions of airborne radionuclides amount to 1.3 percent or the !GC. It is expected that airborne activity conc'ntrations actually observed during plant operation will be s:aall f ractions at the above calculated values in occupied areas.
12.4.3 Occurational Dose Estinates An estimate ot annual man-rem doses associated with plant operation including normal ooeration and maintenance activity has been made trem data contained in the Atomic Industrial Fortrn's !!ational Environmental Studies Project Report entitled "Co:rpilation and Analysis of Data on Occupational Radiation E:mosure Excerienced at Operatina Muclear Pcwer Reactors," dated September, 1974. These estimtes are su=marized in Table 12.4.3-1 and are discussed in this section.
12.4-1
ShTS-1 FS7J1 The man-rem estimate is not a goal but rather an estimate of exposure. The goal is to reduce the exposure associated with each phase of plant operation and maintenance to the minimur, level consistent with as low as practicable considerations for accomplishing each task.
To achieve this goal, the plant desicn includes numerous significant design improvements over present generation plants to reduce occupational exposures.
12.4-2
ENPS-1 FSAR TAbl .I'_ 12 . 4 . 3 - 1 EET3 flisTF S GI' TsM!UT,I, l>OSI:S (1) kork tirca Productive Tank E::posure 1: ate h'ork Time I :c n- l'xt'onurc ( 2 )
(Sun-Tank) _
_(g.r e:r/ hr) (hr/ man) , I'cwer ,{gon-rea)
LI.D Cl.ange out (I t.C 'i s c h n ) 123 11 2 2.9 (techanien) 123 78 12 20.9 CRD aepair
( . echanien)
. 40 78 3 9.6 IS1-Dr y'.iell l'iping (f .echan i en) 150 48 16 31.6 (SDT Techu; 150 11.1 2 3.6 Retueling (I.ech a n i cs) 20 130 5 13.4
!< e ci r e . l un.p I:aint .
(ILC Te:chu) 205 0.4 1 0.09 (I'.t c ha ni e n) 75 70 12 21.5 1:adwinte Synt. ? lai n t .
(I'tchanico)
. 150 75 12 36.7 Main Stt>arr laliel Valve IMant.
(ISC Techs) 75 6 1 0.54 (Mechanien) 100 40 0 17.3 121d ". ! :u l n t .
(ILC Techs) 100 12 4 6.1 1sIV fiaint.
(f.l t e t rici a r.n ) 100 2 2 0.62 (Mechanien) 2b0 25 12 30.9 Unubter insp*-et.lon/
Repair p.echanico) 120 20 t2 9.4 hWCS liaint.
(Li ectr i cic.n c) 250 5 2 2.6 (Mechanica) 200 13 3 8.2 hl!R Systt:1. Maint.
(F.t chanics) 86 24 2 5.0 C[D
,4Li D M~ 7 A {[bL x -
1 oi 2
. ShPS-1 FSAR TitBLE 12.4.3-1 (Cont'd) ,
}.'ol }; l.1 est Titsks D:porure hate Productive j.an- 12:t onu t e ( 2 )
(Su t t 'l a : l.) _ g e.m/hr) Y()[k.bl[0 _l'ows r (tran-J on Turliine V;or k (fleclutnics)
(UDY Techa) 2 180 6 3.1 Speciiil F.aint.
(La t,or er n ) 100 125 20 53.8 (Mechanic:;) 300 125 34 11 3 . 3 I!ealt h l'hysics - - -
65 l'e r Conn (.1 outage b Ruutine - - -
437 Opera tion:,
Rout i n e Maint. .
(Mechanics) - - -
146 TOTAL 583 1
Data obta.inad frc:a Atcrde Industrial Form, Inc.; " Study of the Effectc of Reduced Occupationnl hndiction Enocure I,in.i us on the Co:r.ercial Lu- ,
g, c: cur Power Industry"; July, 1974 ictir tna r.aven cro taced on a 3 y) rea/qnrur occupational doco linit.
An $ ncreac/ccreca factor, discucced in the AIF Study, ~is included.
(3) Although the catinate for the selected tasks given is 427 nan-rcn, a total estinate of 437 man-rer, is eatinatcd in the AIF Study for all outages and routino operatiens.
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SNPS-1 FSAR Request 420.39:
Your description in Sections I.A. and I.B. of your " Fire Hazards Analysis Report" does not provide adequate inrormation on your fire protection organization for us to complete our review.
Thererore, pleaue provice the following information:
(1) You state that prior to fuel loading the Vice President, Project Management will have overall responsibility for the fire protection program and that after ruel loading the responsibility will transter to the Vice President, Operations. Describe the orfsite position that has direct responsibility tor implementing the provisions of the rire protection program. This should include the responsibility for formulation, implementation, and assessment or the effectiveness of the fire protection program and activities such as fire drills and training conducted by the fire brigade and plant pe sonnel.
(2) Describe who has overall responsibility for the rire protection program at the racility. Descrine any delegation or this responsibility for the rire prote ction program such as maintenance or 11re protection equipment and systems, testing or rire protection equipment, fire safety inspections, tire fighting procedures, and tire drills.
(3) Describe the authority of your fire trigade leader relative to that of your Watch Engineers.
Response
(1) The Manager of Electric Produc% ion, Nuclear, who reports to the Vice President, Operations, is responsible ror the overall tor mulation , administration, anc implementation or the Shoreham Fire Protection Pragram oII-site. He provides technical direction to the Plant Manager to whom he has delegated tnese responsibilities. In aadition, as Director of the buclear Review Board (NRB) , the Manager or Electric Production, Nuclear, assures that independent review and audits of the program are conducted under tne cognizance or the NkB in accordance with the Shoreham Tecnnical Specifications at the station. This independent review and audit assesses the efrectiveness of the Shorenam Fire Protection Pragram, training of the station Fire Brigade, and the rire drills conducted at the station. Refer to Section 13.4.2.2 for a description of the NRB.
(2) The Plant Manager has overall on-site responsibility for formulating, implementing, administering, and pe riodically assessing the errectiveness of the Fire Protection Program.
His responslullities are delegated as rollows:
420-39
SUPS-1 FSAR Fire Protection Program Manacer (Maintenance Engineer)
(a) Periodic plant inspections for nousekeeping, accumulation or combustibles, integrity or 11re barriers, tire retardant coatings, and fire protection systemn.
(b) Testing and maintenance of fire protection systems.
(c) Review and evaluation of proposed work activities to identify potential fire hazards, as appropriate to the work activities.
(d) Indoctrination or all plant personnel in the applicable portions of the Fire Protection Program. '
(e) Developing and implementing a Fire Fighting Training Program and organizing a Fire Brigade.
(f) Instructing personnel in the proper handling of accidental events such as leaks or spills of 11ammanle materials. 1 I
Operatina Quality Assurance (OQA) Encineer
[
The Plant Manager assigns to the OQA Engineer the responsibility to apply the appropriate LILCO QA Program requirements to assure that the necessary level or rire protection is maintained during operations. This includes the development and impleLentation of OVA procedures for the review, audit, surveillance, and periodic assessment or the {
Fire Protection Program. j Trainino Coordinator '
f Responsible for implementing training requirements.
Station Fire Chiet k i
The Plant Manager designates a btation Fire Chiet (usually f the Maintenance Foreman) for directing rire righting 4 activities. 2 (3) The on-shitt Watch Engineer has direct responsinility f or the sate operation 01 the plant. However, during orf-hours, he etrectively acts as the Plant Manager. In addition, ne may also tunction As the Fire Brigade lead 3r until relieved or tnose duties by the designated Station Fire Chiet (usually the Maintenance Foreman) .
- I All of the above responsibilities are described in detail in '
the Standard Technical Spec 1rications anc Station Procedures I (SP) listed below: j t
b 420-39a
SNPS-1 FSAR (a) Technical Specification Section 6.0 - Administration Control (b) SP 39.001.01 - Organization and Acninistration of Fire Protection Program (c) SP 39.001.02 - Fire Brigade Organization, hesponse and Drill (d) SP 39.001.03 - Fire Protection Program Training c
[
420-39b
SUPS-1 FSAR Roquest 420.40: ,
The response in Section C of your May 5, 1977 submittal does not indicate whether the QA program for fire protection is under the management control of the QA organization. Thls control consists of (1) formulating and/or veritying that the fire protection QA ,
program incorporates suitable requirements and is acceptable to the management responsible for tire protection and (2) veri 1ying ,
the effectiveness of the QA program for fire protection through revieu, surveillance, and audits. Performance or other QA program functions for meeting the fire protection program requirements may Le performed by personnel outside or the QA organization. The QA program for fire protection should be part ,
of the overall plant QA program. These QA criteria apply to '
those items within the scope of the tire protection program, sucn as fire protection systems, emergency lighting, communication and emergency brea thing apparatus as well as the tire protection i requirements of applicable safety-related equipment. i Re s r>on s e . .
t The Operational Quality Assurance Engineer has direct responsibility for assuring implementation or the LILCO QA i Program at the station, including its appropriate application to ,
the fire protection program. He is responsiDle for assuring the establishment o1 design, procurement, installation, testing, and administrative controls for the fire protection program during operation phase activities. He assures that the program is being
- effectively implemented by means of inspections, surveillance, and scheduled audits. He assures and verliies that the results of inspections, surveillance, and audits are reported to :
cognizant management personnel in accordance with approved procedures.
l The Quality Assurance Manager is ultimately responsible for the :
development and implementation or the overall Quality Assurance Program during design, construction, preoperational testing, operataon, and major modifications as discussed in Section :
17.2.1.
l f
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1 420-40 l
REQUEST 420.41 Although you have addressed the ten specific quality assurance criteria in Branch Technical Position APCSB 9.5-1, we find that your response does not describe sufficient detail for thcse criteria. In order for the OAB to fully evaluate these criteria, additional detailed description is necessary. Examples of the detail we would expect Long Island 'ighting Company to consider are provided in Attachment 6 of Mr. Vassallo's letter of August 29, 1977. If, h7 wever, you choose not to provide this detail, you may apply the same controls to each criterion that are commensurate with the controls described in your operational QA progra.T (after acceptance by NRC). These controls would apply to the remaining construction activities for Unit 1 and for the operations phase of Unit No. 1. If you s' elect this method, a statement to this effect would be adequate for our review of the fire protection QA program.
RESPONSE
The same controls will be applied to each criteria commensurate with the controln described in the QA program during the operational pl.are with exceptions as listed in the Fire Hazards Analysis Report, Section 1.C. Refer to FSAR Section 17.2, " Quality Assurance (QA) During the Operations Phase" for additional details.
420-41
A SER OPEN ITEM 1 Concern was expressed that some quantity of water could be carried into the intake structure through the seaward facing ventilation openings on the building's roof. The condition leading to this event, as postulated by the NRC Staff, is a combined surge and wave runup reaching 43 feet MLW. Accord-ingly, the intake structure has been evaluated for wave runup, breaking wave forces, and uplift pressures from an 8 foot wave, concurrent with a stillwater '.evel of 22 feet MLW, resulting in a 21 foot runup on the seaward face of the intake structure to elevation 43 feet MLW The results of the analyses performed indicate that this condition results in no forces on the intake structure more severe than those already established, and for which it has been designed. Therefore, the structural integrity is demonstrated. The elevation of the intake structure roof is 40 feet MLU. The vents are located on the landward end of the roof, with the seaward face of the vents 20 feet from the seaward face of the structure. The cross section of a typical vent is shown on Figure 1. As can be seen from the figure, wind driven spray from the wave runup, in order tc enter the pumphouse, must enter the vent and pass over a 4 foot 10 inch wall and through the screened opening and louvered damper.
We believe that with this design, water entering the pumphouse would be insufficient to present a hazard to safety-related systems and components. Nevertheless, to further ensure no hazard exists, adaitional protection has been provided against entry of water through the roof cop vents. This protection is in the form of 4 feet 4 inch high concrete walls located seaward of the vents as shown on Figure 2.
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SER Open Item 11 - Target Rock SRV Qualification Test Temperature Profile Letter SNRC-357 provided a summary of the Target Rock safety /
relief valve environmental test results. The last test interval (650/5521 minutes) was run at temperatures between 290 and 301 F.
The NRC has questioned the equivalency of this test interval relative to the FSAR-stated long term post-LOCA condition of 150 F for 99 days.
It is normal industry practice to conduct accelerated life time / temperature interpolation tests by increasing the test temperature parameter. Chemistry reference books and IEEE 117-1974, Page 13, Table 1, subscribe to the "10 degree rule" for deterioration of non-metallic equipment components such as insulation and seals. Generally, this rule states that for each 10 C rise in test temperature, con.ponen t life is reduced by 50 percent. On this basis, the final test interval of the Target Rock test procedure was significantly conservative.
.