ML19318B557
| ML19318B557 | |
| Person / Time | |
|---|---|
| Site: | Brunswick  |
| Issue date: | 06/11/1980 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML19318B555 | List: |
| References | |
| NUDOCS 8006270039 | |
| Download: ML19318B557 (5) | |
Text
_ _ _ _ _ - _ _ _ _ _ _
+$puutuqhg UNITED STATES
/7 C'
NUCLEAR REGULATORY COMMISSION 3
y WASHINGTON, D, C 20886
\\.....,o8 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION SUPPORTING AMENDMENT NO. 28 TO FACILITY LICENSE N0. DPR-71 AND AMENDMENT NO. 51 TO FACILITY LICENSE NO. DPR-62 CAROLINA POWER & LIGHT COMPANY BRUNSWICK STEAM ELECTRIC PLANT UNIT NOS. 1 AND 2 DOCKET NOS. 50-325 AND 50-324 A.
Brunswick Steam Electric Plant, Unit No. 2, Fuel Cycle 4 -
Reload Application By letter dated February 20, 1980, as su'.plemented April ll,1930.
Carolina Power and Light Company (the licensee or CP&L), requested amendments to Facility Operating License No. DPR-62. The proposed changes relate to the replacement of 132 fuel assemblies constitut-ing refueling of the core for fourth cycle operation at power levels up to 2436 Mwt (100% power) for Brunswick Steam Electric Plant Unit No. 2 (BSEP2).
CP&L had requested credit for the end-of-cycle recirculation pump trip (E0C-RPT) feature in the previous Cycle 3 reload proposal. As indicated in the Safety Evaluation for Amendment 48 to DPR-62 dated May 2,1979, the staff had reservations about the design implementa-tion of the E0C-RPT system. Our concerns were associated with veri-vication testing and electrical design interfaces. By letter dated May 21,1980, the licensee provided information on the BSEP2 E0C-RPT I
verification testing. The EOC-RPT electrical design interface is discussed in section C of this SER.
Corrective Action for Reload MCPR Error
]
By letter dated April 23, 1980, the licensee provided a summary of corrective action taken to avoid recurrence of an error that i
occurred during licensing of the previous SEP Unit 2 reload. The error involved incorrect computer input data that resulted in non-conservative minimum critical power ration (MCPR) limits foi the latter part of the operating cycle.
It appears that the corrective action taken will be adequate to preclude similar errors in future l
reload proposals.
The staff was assisted in the Safety Evaluation of the Brunswick 2 Reload-3 licensing analysis by our technical consultant, Brookhaven National Laboratory (BNL). The following evaluation was sutnitted by BNL on May 27, 1980.
l 800 6 270 OM
SAFETY EVALUATION OF BRUNSWICK 2 RELOAD-3 LICENSING ANALYSIS 1.
INTRODUCTION A safety evaluation has been carried out for Carolina Power and Light Com-This pany's Brunswick Steam Electric Plant Unit 2 Reload 3 (BSEP-2 R-3).
plant is a BWR-4 which contains 560 fuel assenblies. The Reload 3, or Cycle 4 core is expected to include 132 fresh P8 x 8R type assemblies, approximately 24% of the core.
In support of its application for a reload license to operate the BSEP-2 R-3, Carolina Power and Light (CP&L) has submitted along with a letter to the NRC, a proposed set of revisions to the Technical Specifications as well as l
the Supplemental Reload Licensing Submitta12 which includes credit for the end-of-cycle recirculation pump trip (E0C-RPT) system. This supplemental sub-mittal is also referred to as the PPT-analysis submittal.
In a subsequent transmittal to the NRC3 CP&L submitted a new set of revisions to the Tech-nical Specifications and a revised version of the Supplemental R31oad Licens-ing Submittal.4 The revisions to the technical specifications refer to op-4 does eration with the RPT system whereas the revised Supplemental Subr.ittal not include credit in thermal margins for the EOC-RPT. MCPR Ifmits in the non-RPT analysis are more restrictive. These more restrictive Operating Limit MCPR's (OLMCPR) are to be observed if the EOC-RPT system becones inoperable.
Reference 4 is also known as the non-RPT analysis submittal.
With the exception of the sections of this document where MCPR limits and E0C-RPT are discussed, the evaluation applies to both RPT and non-RPT cases.
. Reference 5 provides values for plant-specific data, such as steady state operating pressure, core flow, safety and safety-relief value setpoints and other design parameters. Additional plant and cycle specific data are pro-vided in the Supplemental Reload Transmittals 2,4 These latter documents follow the outline presented in Appendix A_ of Reference 5.
The Licensee has agreed to supply the Commission with comparisons of measured to calculated pump coastdown data, which will provide the basis for our consideration of the RPT feature. Our acceptance of the analysis for the RPT operation depends on the favorable outcome of that comparison.
Our evaluation of the Brunswick 2 Reload 3 is limited to the items dis-cussed in the following sections.
2.
EVALUATION 2.1 Nuclear Characteristics The 132 fresh P8 x 8R assemblies are designated as P8DRB265H. The Cycle 4 core will have 428 previously exposed Reload 1 and 2 8 x 8 and 7 x 7 assem-blies, including 152 bundles from the initial core. The breakdown of these bundle types is listed in Re 7erence 2.
Results of calculations presented in Section 4 of Reference 2 show that throughout the cycle both the control rod system and the standby liquid control system will have adequate shutdown capa-l bility under the most reactive conditions of the core.
2.2 Thermal Hydraulics 2.2.1 Fuel Cladding Integrity Safety Limit-MCPR According to Reference 5, the safety limit MCPR (SLMCPR) for BWR-4 cores which refuel with 8 x 8R and P8 x 8R is 1.07.
This limit implies that during J
l
. a transient characterized by an MCPR of 1.07, 99.9% of the fuel rods in the core are expected to avoid boiling transition.
We note that the use of pre-pressurized fuel rods in the 132 fresh bundles planned for loading in the Cy-cle 4 core will have the effect of slightly reducing fuel temperatures during power operation. This effect will result in a small reduction in the local Doppler feedback effect on pin power peaking. The resulting difference in the local peaking between 8 x 8R and P8 x 8R assemblies is reported to be in-significant (References 6 and 7).
Furthent; ore, higher peaking in the prepres-surized (P8 x 8R) retrofit assemblies would tend to reduce the flatness of it.-
trabundle peaking and since decreased peaking (flatter power distribution) re-suits in more rods in boiling transition in the GETAB analysis, the use of the 8 x 8R R-factor distribution for P8 x 8R reloads is considered conservative.
The staff has found this approach in arriving at the statistical safety limit, originally derived for 8 x 8R BWR reloads, to be equally acceptable for the P8 x 8R reloads. (Reference 9) 2.2.2 Operating Limit MCPR To establish the OLMCPR for the Brunswick 2 Cycle 4 operation, two sepa-rate transient analyses were performed with the REDv Code:
(a) for operation with the RPT system 2, and (b) for operation without the RPT system.4 As mentioned earlier, OLMCPR's for the RPT operation are less restrictive. Oper-ating limit MCPR's for each type of fuel in the Cycle 4 core are given in Reference 2 for the RPT operation. Additional confirmatory analyses have been performed with the ODYN code for thir. case. For operation without the RPT, thermal margins are reduced. OLMCPR's for non-RPT operation are given in Reference 4.
. Operating MCPR limits for the non-RPT case are found acceptable.
- However, for the RPT operation, the OLMPCPR for the 8 x 8R fuel must reflect the penalty associated with the use of the ODYN code as discussed in Section 2.2.2.1 and in accordance with Reference 9.
The raw initial CPR (ICPRnew) is obtained is follows :
STEP 1.
ODYN-Calculated ACPR:
ACPR = 0.14 c
STEP 2..
Calculated ICPR; ICPR = 1.07 + 0.14 = 1.21 c
where the GETAB limit i..1.07 ACPRc 0.14 0 116
=
ICPR g
c STEP 3.
ACPR" 0.116 + 0.044 0.160
=
=
ICPRnew 1.07 Since ICPR
- ACPR
=
new new or (1;00 - 0.16) ICPR 1.07
==
new ICPR new Hence, the new initial OLMCPR for the 8 x 8R fuel is 1.27.
e
. 2.2.2.1 Transient Analysis Methods The generic i 9thods used for these calculations, including cycle-independent initia', conditions and transient input parameters, are described in Reference 5.
The staff evaluation, included as Appendix C o.' Reference 5, contains the acceptance of the cycle-independent values. The evaluation of the transient analysis methods, appears in Appendix C of Reference 3.
Suppl e-mentary cycle-independent initial conditions and transient input parameters used in the transient analyses for both RPT and non-RPT cases appear in the tables in Sections 6 and 7 of References 2 and 4.
The evaluation of the meth-ods used to develop these supplementary input values is also included in Ref-1 erence 10.
For plants which utilize the RPT system, additional analysis or justifi-cation has been required, since it has been concluded that the current methods do not adequately model the RPT phenomena. Therefore, the licensee has pre-sented an ODYN transient analysis with RPT. ODYN is an improved transient an-alysis code which has been used in the past to model RPT phenomena (TVA Ref-erence 8). However, due to uncertainties in the ODYN code, a penalty of 0.044 in ( CPR/ICPR) must be applied to the licensing calculations, as described in Reference 9.
2.2.2.2 Transient Analysis Results The transients evaluated were the limiting pressurization and power ir.-
crease transients (generator load rejection without bypass, the feedwater controller failure and loss of feedwater heating), and the control rod with-drawal error. The analysis results of the fuel loading error have been in-co.'porated in the specification of the operating limit MCPR per Reference 5 i
l
. (see Section 2.3.3).
Initial conditions and transient input parameters as specified in Sections 6 and 7 of both References 2 and 4 were assumed.
The results of these analyses are outlined in Sections 9 and 10 of both References 2 and 4.
It is acceptable if fuel specific operating limits are established for prepressurized fuel (Reference 10). On this basis, the trans-ient analysis results are acceptable for use in the evaluation of the opera-ting limit MCPR.
2.3 Accident Analysis 2.3.1 ECCS Appendix K Analysis In the safety evaluation of Reference 5, it was concluded that "the con-tinued application of the present GE ECCS-LOCA (" Appendix K") models to the 8 x 8 retrofit reload fuel is generically acceptable" and in the Reference 10 evaluation that conclusion was extended to prepressurized fuel. On these bases, the proposed MAPLHGR limits for the new prepressurized fuel are accept-able.
2.3.2 Control Rod Drop Accident The significant parameters in the rod drop analysis satisfy the require-ments for the bounding analyses described in Reference 5.
Therefore, the results of this analysis are well below the acceptance criterion of 280 calo-ries per gram.
2.3.3 Fuel Loading Error The GE method for analysis of misoriented and misloaded bundles has been reviewed and approved by the staff and is part of the Reference 5 methodology.
Potential fuel loading errors involving misoriented bundles and bundics loaded
. into incorrect positions have been analyzed by this methodology and the ie-sults have been incorporated into the specificatian for operating limit MCPR.
This assures that SLMCPR is not violated for any potential error in either orientation or loading of a bundle.
2.3.4 Overpressure Analysis The overpressure analysis for the MSIV closure with high flux scram, which is the limiting overpressure event, has been performed in accordance with the requirements of Reference 10. We agree that there is suffi.ient margin be-tween the peak calculated vessel pressure and the design limit pressure.
Therefore, the limiting overpressure event as analyzed by the licensee is con-sidered acceptable.
2.5 Technical Specifications The Technical Specifications have been changed to include specifications associated with the new, prepressurized type bundles.
Reference 10 contains the revised limiting conditions for operation as well as the corresponding surveillance requirements, regarding the Average Planar Lt ear Heat Generation Rates ( APLHGR's), the APRM and Rod Blnck Monitor setpoints, and Linear Heat f
Generation Rates (LHGR's) applicable to the Reload 3 core operation for the l
RPT case. Technical Specifications for OLMCPR's for the RPT operation must be consistent with the values established in this evaluation for the 8 x 8R fuel.
Technical Specifications changes reflecting limiting conditions for operation and surveillance requirements as well as Ob.CPR's resulting from the introduc-l tion of the new type of bundles for the non-RPT case have been reviewed and f
foJnd acceptable.
1
. 2.6 Densification Power Spiking It is acceptablell to remove the 8 x 8, 8 x 8R and P8 x 8R spiking pen-alty factor from the Technical Specification of those BWR's for which it can be demonstrated that the predicted worst case maximum transient LHGR's, when augmented by the power spike penalty, do not violate the exposure-dependent safety limit LHGR's. The Brunswick 2 plant meets the above criterion.
Sec-tion 10, Rod Withdrawal Error and Appendix E Linear Heat Generation Rate for Bundle Loading Error, of Reference 2 and 4 include the densification effect in the reported LHGR value for all 8 x 8 type assemblies. On the basis of these data, we find that the licensee meets the requirements on the densification power spiking.
2.7 Recirculation Pump Trip Feature In Appendix C of Reference 2, a qualitative description of the Recircu-lation Pump Trip (RPT) feature is given. The purpose of the RPT systsi is to mitigate core-wide pressurization transients by a rapid reduction in core flow and an increase in the core void content thereby reducing the peak transient power and heat flux.
The preposed Technical Specifications indicate that the RPT system shall be demonstrated operable on a monthly basis.
In evaluating the OLMCPR's for the RPT operation, it has been assumed tnat the RPT system is operable. Questions related to the hardware and reliability of the RPT system are outside the scope of this review and therefore are not being addressed in this work.
Our acceptance of the analysis for the RPT operation is based on the re-ceipt of the additional infomation, agreed to by CP&L, leading to favorable comparison between measured and assumed pump coastdown data.
l i
.g.
REFERENCES 1.
Letter, E. E. Utley (Carolina Power and Light Company) to T. A. Ippolito (USNRC) February 20, 1980.
2.
J. L. Rash, " Supplemental Reload Licensing Submittal for' Brunswick Steam Electric Plant Unit 2 Reload 3,"
NED0-24235, General Electric Company, Janua ry,1980.
3.
Letter, E. E. Utley (Carolina Power and Light Company) to T. A. Ippolito (USNRC), April 11, 1980.
4.
J. L. Rash, " Supplemental Reload Licensing Submittal for Brunswick Steam Electric Plant Unit 2 Reload 3, NED0-24235 Revision 1, General Electric Company, March 1980.
5.
" General Electric Boiling Water Reactor Generic Reload Application,"
NEDE-24011-P-A, August, 1979.
6.
General Electric letter (E. Fuller) to USNRC (0. Parr), June 8,1978.
7.
General Electric letter (E. Fuller) to USNRC (0. Parr), August 14, 1978.
8.
Letter, T. A. Ippolito (USNRC) to H. G. Parris (TVA) dated February 8, 1979 and enclosed SER.
9.
Letter, R. P. Denise (USNRC) to G. G. Sherwood (GE) and enclosure, dated January 23, 1980.
- 10. Letter, T. A. Ippolito (USNRC) to R. Gridley (G.E.), April 16, 1979 and enclosed Safety Evaluation Report.
11.
Revised Technical Specifications for Carolina Power and Light Brunswick Steam Electric Plant Unit 2.
( Attached to Reference 1)
. 3.0 Conclusions Based on our review of the consultant's Safety Evaluation and our own examination of the recirculation pump coast-down characteristics provided by CP&L in their letter dated May 21, 1980, we find the proposed operation in Cycle 4 allowing credit for E0C-RPT to be accepta ble. The May 21, 1980 letter also committed to perform the same series of physics tests that were judged acceptable during the previous refueling.
We find this startup test program acceptable for Cycle 4 operation. As indicated in the BNL evaluation, the Technical Specification Operating Limit MCPR for 8x8R and P 8x8R fuel has been revised to include the required ODYN penalty for E0C-RPT operation.
B.
Brunswick Steam Electric Plant, Unit Nos.1 and 2 Degraded Grid Voltage Protection
1.0 INTRODUCTION
In response to +.he NRC's generic letter of June 3,1977, CP&L proposed design modifications and changes to the Technical Specifications in accordance with the criteria and staff positions contained in our letter.
2.0 MODIFICATIONS The following modifications were proposed in response to the generic letter of June 3,1977:
a)
Installing second level undervoltage relays, three on each of the four 4160v Class lE bus with a drop out setting at 89.5% of nominal bus voltage and a 10 second time delay in coincident trip logic (2 out of 3) for degraded grid voltage protection.
b) Blocking the load shedding feature on the 4160v Class lE buses when the diesel generators are supplying to these buses, and automatically reinstating this feature when the diesel generator breakers are tripped.
3.0 EVALUATION The acceptability of the BSEP Units 1 & 2 Degraded Grid Voltage Protection modifications was reviewed for the staff by our technical consul tant, EG&G Idaho, as part of the Selected Electrical, Instru-mentation, and-Control Systems Issues Program. The Lo:is fo
- ac ept-ance is documented in " Technical Evaluation Report of tne De6'aded Grid Protection for Class lE Power Systems at the Brunswit; ctec Electric Plant, Unit Nos.1 and 2" dated February
.'d0 whicn is inct +-
porated herein by reference.
e
4.0 CONCLUSION
S Based on our review of the referenced consultant's Technical Evaluation Report, we agree with their findings that the modifications will protect the safety related equipment from a sustained degraded voltage of the offsite power source and the design bas'es meet the Staff's Positions. Also, we agree with the consultant that the proposed changes to the Technical Specifications adequately address testing of the protection systems and comply with Staff Position 3.
Therefore, we conclude that CP&L's proposed design modifications and changes to the Technical Specifications are acceptable.
C.
Brunswick Steam Electric Plant, Units 1 and'2 End-of-Cycle Recirculation
~
Pump Trip 1.0 INTRODUCTICN By letter dated February 2,1979, as supplemented March 21, 1979, April 13,1979, April 27,1979, May 1,1979, May 29,1979, February 5, 1980, and April 11, 1980, the licensee provided information regarding the prompt recirculation pump trip feature. The letters dated April 13,1979, April 27,1979, and April 11, 1980 proposed Technical Specifications for the E0C-RPT.
2.0 DISCUSSION The design philosophy for the E0C-RPT system is described in General Electric's (GE) report NED0-24119 " Basis for Installation of Recircula-tion Pump Trip System," Browns Ferry Nuclear Plant, April 1978. The application of the E0C-RPT to Brunswick Unit 2 was described in GE report NED0-24587, SUPPLEMENTAL RELOAD LICENSING SUBMITTAL FOR BRUNSWICK STEAM ELECTRIC PLANT UNIT 2 RELOAD 2 (RECIRCULATION PUMP TRIP FEATURE) dated January 1979 and GE report NED0-24179 Revision 1 (same title) dated March,1979.
3.0 EVALUATI_0N The acceptability of the BSEP 2 E0C-RPT feature was reviewed for the staff by our technical consultant, : Lawrence Livermore Laboratory, as part of the Selected Electrical, Instrumentation, and Control Systems Issues Program.
The basis for acceptance is documented in
" Technical Evaluation of the End-of-Cycle Recirculation Pump Trip for Brunswick Steam Electric Plant Unit No. 2" dated April 1980 which is incorporated herein oy reference.
4.0 CONCLUSION
S Based on our review of the referenced consultant's Technical Evalua-tion Report, we conclude that the E0C-RPT feature for the BSEP meets applicable design criteria. We find the proposed design acceptable for both units 1 and 2.
The proposed Technical Specification changes for BSEP Unit 2 to incorporate the E0C-RPT feature are acceptable, as modified by discussions with the staff.
We have determined through discussions with the licensee, as documented in their subnittal dated May 21,1980, that the BSEP Unit 2 E0C-RPT system response time is less than the response time assumed in the transient analysis for the BSEP Unit 2 reload application. A similar verification will take place prior to authorizing credit for E0C-RPT for BSEP Unit 1.
We, therefore, find the EOC-RPT feature acceptable.
D.
Brunswick Steam Electric Plant, Unit Nos.1 and 2 Reactor Vessel Water Level Instrumentation
1.0 INTRODUCTION
By letter dated January 18, 1979, the licensee requested revisions to the Technical Specifications for BSEP Units 1 and 2 concerning operability requirements for reactor vessel water level low and low-low instrumentation. These changes were requested to allow the performance of maintenance and modification work on the reactor vessel and ancillary systems during cold shutdown and refueling.
2.0 DISCUSSION The Commission issued Amendment No.18 to Facility License No. DPR-71 for BSEP Unit No. 1 on January 19, 1979 which granted a one-time Special Test Exception to allow lowering the reactor vessel water level for extended maintenance during the then current refueling i
outage. The staff indicated that prior to authorizing a permanent change to the Technical Speci#ications providing this flexibility, further review would be required to judge the acceptability of this change on a generic basis. Of specific concern was the possibility for inadvertently draining the reactor vessel.
3.0 EVALUATION The staff's continuing review revealed no creditible mechanism for draining the reactor vessel inadvertently while in modes 4 or 5.
In particular, there were no events which could be postulated that would overpressurize or rupture low pressure piping systems and lead to draining the vessel.
Even considering a non-mechanistic failure resulting in loss of vessel inventory, a level decrease (to L3) will automatically initiate LPCI to maintain vessel inventory, since the low-low-low level instruments are not affected by this change.
Furthermore, water level. is required to be monitored periodically
. during refueling. As a result of this determination, the GE Standard Technical Specifications were revised to remove the low and low-low water level instrumentation operability requirements with cold shut-down.or refueling conditions in effect.
4.0 CONCLUSION
We conclude that there is no safety related function for low and low-low reactor vessel water level instrumentation in operational condi-tions 4 or 5.
Revising the BSEP Technical Specifications to be consistent with the now current GE Standard Technical Specifications is an acceptable alternative to providing a Special Test Exception, as was originally requested. This modification has been discussed and agreed with by the licensee.
E.
Brunswick Steam Electric Plant, Unit Nos.1 and 2 Corporate Organizational Changes
1.0 INTRODUCTION
By letter dated November 7,1979, the licensee requested a revision to the Technical Specifications for BSEP Units 1 and 2.
The changes would reflect administrative corporate organizational changes which became effective on November 3,1979.
2.0 DISCUSSION The licensee had proposed a corporate reorganization on August 14, 1 979. The November 7,1979 submittal superceded the previous sub-mittal which had not been reviewed by the staff.
3.0 EVALUATION The proposed changes were found to be acceptable.
However, our i
review identified the lack of specific provisions to formalize l
~
1 the requirement that the Corporate Nuclear Safety & Quality Assurance Audit Section review records of the Plant Nuclear Safety Committee activities.
During telephone discussions on this subject, the i
licensee agreed to accept our proposed Technical Specifications for this requirement.
4.0 CONCLUSION
Based on our review of the licensee's subnittal and the agreed upon corrective action for the one identified discrepancy, we find the proposed change acceptable.
a o
6-F.
Brunswick Steam Electric Plant, Units 1 and 2 RHR Service Water Pump Discharge Pressure
1.0 INTRODUCTION
By letter dated December 31, 1979, the licensee requested a revision to the Technical Specifications for BSEP Units 1 and 2.
The changes would revise the operability test requirements for the Residual Heat Removal (RHR) service water pumps.
2.0 DISCUSSION The licensee reported that an error in the data sheet used in the calibration of the flow instrument during the original startup test resulted in an incorrect discharge pressure requirement in the Technical Specifications.
The correct value has been proposed in 9
terms of differential pressure across the pump with a minimum allowable suction pressure and flow rate.
In effect, the minimum discharge pressure has been decreased from 300 psig to 252 psig at 4000 gpm.
3.0 EVALUATION The importance of the service water pump discharge pressure is an equipment specification which requires that the pumps maintain a minimum differential pressure of 20 psid from the tube to the shell side of the RHR heat exchangers, thereby preventing reactor water leakage into the service water system.
Plant procedures permit shutdown cooling initiation at a reactor vessel pressure of 125 psig or below; thus, sufficient service water pressure is available to maintain the heat exchanger differential pressure during the shutdown cooling mode of operation.
In the worst case accident condition (steam condensing mode) the shell (reactor) side of the RHR heat exchanger could experience pressures as high as 183 psig.
With the tube (service water) side of the heat exchanger maintained at a minimum of 252 psig, sufficient service water pressure is available to maintain the heat exchanger differential pressure during the accident mode of operation.
4.0 CONCLUSION
The proposed change to the residual heat removal service water subsystem operability surveillance requirement is conservative.
The change establishes a minimum allowable suction pressure require-ment consistent with ASME Section XI, and provides for a more complete characterization of the required pump performance. We find the proposed change ' acceptable.
- G.
Brunswick Steam Electric Plant, Unit Nos. I and 2 Revision of Environmental Technical Specifications
1.0 INTRODUCTION
By letter dated December 27, 1976, the licensee proposed a change to the Environmental Technical Specifications (ETS) for BSEP Units 1 and 2.
The change would modify the reporting requirements to make them more responsive to plant-associated occurrences.
2.0 DISCUSSION The current reporting requirements for Non-Routine radiological reports (30 day) do not differentiate between increases attributable to the operation of the plant and increases due to other events, such as atmospheric nuclear weapons testing.
It is possible to make this distinction using isotopic analysis methods, with a reasonable degree of accuracy, given that the analyst is aware of off-site events and plant occurrences that could be responsible for any identified radionuclides.
3.0 EVALUATION The licensee is required to provide a separate annual environmental radiological report including survey and sample sunmaries, and environmental data analyses.
Radiological environmental monitoring increases not attributable to plant operation should be reported in the anntal report. We concur that it is not appropriate to report radiological environmental monitoring increases determined to be unrelated to plant operation in Non-Routine reports.
4.0 CONCLUSION
We find the proposed ETS change to be acceptable.
H.
BSEP Unit No. 2 Hydraulic Snubbers
1.0 INTRODUCTION
By letter dated May 27, 1980, the licensee requested a revision to the listing of safety related hydraulic snubbers for Brun: wick Unit 2.
2.0 DESCRIPTION
The proposed revision was the result of a plant modification which identified four (4) hydraulic snubbers that could be replaced with rigid restraints.
It is proposed that the following snubbers on reactor vessel instrument lines be removed from Table 3.7.5-1:
2PS-3558 2PS-3561 l
2PS-3562 2PS-3570
, 3.0 EVALUATION The seismic analysis supporting the proposed snubber deletion was performed using methods developed under IE Bulletin 79-07 by CP&L.
The analysis techniques were accepted by NRC on October 17, 1979.
The revision of Table 3.7.5-1 based on the seismic reanalysis using the approved methods is acceptable.
4.0 CONCLUSION
The proposed revision to Table 3.7.5-1 is approved. Future snubber deletions requiring technical specification changes should be proposed in advance to enable staff concurrence prior to initiating modifications.
I.
ENVIRONMENTAL CONSIDERATION We have determined that the amendments do not authorize a change in effluent types or total amounts nor an increase in power level and will not result in any significant environmental impact. Having made this determination, we have further concluded that the amendments involve an action which is insignificant from the standpoint of environmental impact and pursuant to 10 CFR 551.5(d)(4) that an environmental impact statement, negative declaration, or environmental impact appraisal need not be prepared in connection with the issuance of the amendments.
J.
CONCLUSION We have concluded, based on the considerations discussed above, that:
(1)'because the amendments do not involve a significant increase in the probability or consequences of accidents previously considered and do not involve a significant decrease in a safety margin, the amendments do not involve a significant hazards consideration, (2) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, and (3) such activities will be conducted in compliance with the Commission's regulations and the issuance of the amendments will not be inimical to the common defense and security or to the health and safety of the public.
j Dated:
June 11,1980
)
+
- e 4
I m
i SUPPLEMENT NO. 2 1
TO THE FIRE PROTECTION SAFETY EVALUATION REPORT BY THE OFFICE OF NUCLEAR REACTOR REGULATION e
~
U.S. NUCLEAR REGULATORY COMMISSION IN THE MATTER OF CAROLINA POWER & LIGHT COMPANY BRUNSWICK STEAM ELECTRIC PLANT UNIT N05. 1 and 2 DOCKET N05. 50-325 and 50-324 i
DATE:
June 11,1980 L
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Item 1.0, Protection of The following evaluation addresses four areas.
Redundant Safe Shutdown Cabling (greater than five foot separation),
was identified as an incomplete item in our fire protrtion SER supple-Item 2.0, Protection of Redundant Safe Shutdown ment of April 6, 1979.
Cabling (less than five foot separation), addresses a concern identified during our site visit of March 5 and 6,1980.
Item 3.0, Fire Protection Loop Isolation Valve, was identified in an I&E inspection as a deviation.
Item 4.0, Door Frames for Fire Doors - Diesel Generator Building, discusses a proposed change to a comitment identified in our fire protection SER.
i 1.0 Protection of Redundant Safe Shutdown Cabling (greater than five foot separation)
Our SER supplement of April 6,1979, indicated that the cable study)perfonned by CP&L and United Engineers and Cons (UE&C safe shutdown divisions were routed within five feet horizontally of each other.
However, except for high hazard areas, the study assumed that where cables were separated by more than five feet horizontally, these would not be involved in a single fire.
We were concerned that certain situations could exist where a fire could still affect redundant cabling, even if separated by five feet horizontally. Accordingly, our SER supplement of April 6,1979 noted that this issue was still under review.
On March 5 and 6,1980, we visited the site to survey those areas where it appeared that safe shutdown cabling came to within 5 to 15 feet of each other. Areas surveyed were the diesel generator ouilding basement, service water structure basement, and general open areas of both reactor buildings.
Based on our evaluation of these areas, we find that adequate separation and/or fire protection (sprinklers and barriers) have been provided to assure that fires that could be postulated for these areas will not cause loss of redundant safe shutdown cabling. Accordingly, we find that fire protection for safe shutdown cabling separated by more than five feet horizontally satisfies the objectives identified in Section 2.0 of our SER and is therefore acceptable.
n q
. 2.0 Protection of Safe Shutdown Cabling (less than five foot separation)
Our SER supplement of April 6,1979, noted that adequate protective
- easures had been proposed for redundant safe shutdown cabling that was routed within five feet of each other. These protective measures included use of thermal blankets, 3-hour fire enclosures, sprinkler systems, and cable coatings as required to protect from postulated fires. These measures were documented in the cable study report submitted March 30, 1978 and subsequent correspondence from CP&L of March 15,1979, and were to be completed during the 1979 refueling outages.
During our site visit of March 5 and 6,1980, we identified two locations where redundant safe shutdown cabling were routed in close proximity to each other such that a fire in cable trays of one division could cause loss of cables in conduit from the other division.
In each of these areas, it appears that the problems were not identified in the cable study referenced above, but apparently should have been.
The following discusses the concerns in these two areas:
Intake Structure Basement - At one end of this area, cabling in conduit associated with the green division lube water pump for Unit 1 is routed adjacent to and directly over red division cable trays. Aprinkler heads are located in the area. including below the red division trays; however, no thermal barriers are provided.
With the present protection a fire in the red division trays could also damage the green division. cabling in conduit prior to actuation of the sprinkler system, or if the sprinkler system fails to operate.
The potential for a fire to affect this redundant cabling was not identified in the licensee's cable study; although, with the criteria used in the study, this problem should have been identified.
By letter dated April 1,1980, the licensee has proposed to install 1-1/2 hour fire-rated thermal insulation on the green-division conduit associated with the Unit I service water lube pump to protect it from a fire in the red division.
cable trays.
Unit 1 Reactor Building - Elevation 20 Feet - In the southwest area at this elevation, cabling associated with the green division is routed vertically within two feet of a horizontal red division cable tray, and then is routed parallel to the red tray for a distance of approximately 25 feet at an elevation of approximately four feet greater than the cable tray.
The concern is that a fire in the red tray could damage
3-the green division cable in conduit.
No sprinkler or thermal barriers are provided in this location. The green cable is routed to a motor operator associated with the RHR system.
This green division motor is located adjacent to a stack of six red division cable trays; because of this, there is also concern that a fire in these trays could affect operability of this valve, as well as loss of the redundant red division RHR system due to cable involved in the fire. As in the intake structure, these problems should have been identified in the original cable study, but were not.
By letter of April 1,1980, the licensee has proposed to install 1 1/2 hour fire-rated thermal insulation on the green division conduit identified above up to the valve operator, and to provide a sprinkler head to protect the motor operator from a fire in the red division cable trays.
In addition, the red division cable trays are coated with a fire retardant coating where they are in proximity to the valve operators. With the changes described above, we find that protection provided for these two identified areas of concern satisfy the objectives of Section 2.0 of our SER and is therefore accepta ble. We also find the proposed schedule for implementa-tion to be acceptable.
3.0 Fire Protection loop Isolation Yalve As noted in our SER of February 22, 1977, the licensee had proposed to add additional sectionaltzing valves on the yard loop and that the loop would confom with NFPA 24, "Outside Protection."
NFPA 24 requires that all control valves in the fire protection water system' be indicating type except where required by special conditions, and accepted by the authority having jurisdiction.
Part of the fire rotection loop runs across the turbine building and has a curb (key operated) valve installed in a covered pit as a sectional control valve. A bypass controlled by noma 11y open post indicator valves is installed around this valve to afford added flexibility in control of the system. This arrangement is acceptable on the basis of:
(1) general inaccessability of the valve in the pit with a steel plate cover; (2) the valve being key operated, i.e., a hand wheel is not installed on the valve; and (3) periodic surveillance of valve position by the licensee.
4.0 Door Frames for Fire Doors in Diesel Generator Buf1 ding Our SER originally called for all fire door frames to be labeled by a recognized testing laboratory.
The licensee has modified a number of frames for fire door installation in the Diesel Generator Building and provided details in a letter dated February 25, 1980, At '
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. to denonstrate that these modifications make the frames satisfactory f6r three-hoar fire-rated service. We h:ve evaluated details of the modification and visually examined the modifications during a site visit of March 5 and 6,1980.
We find that the fire door frames as modified are satisfactory for the service intended to prevent fire spread through the protected openings in the Diesel Generator Building fire barriers, and are therefore, acceptable.
5.0 Fire Barrier Penetration Technical Specifications Our letter dated February 14, 1980 requested the licensee to revise the Technical Specifications for fire barrier penetrations at BSEP to conform to the current General Electric standard.
Model Technical Speci-fications were provided as guidance.
The licensee's response dated April 22,1980 proposed modified Technical Specifications designed to differentiate between barriers based on size of opening.
The staff has reviewed the licensee's proposal and found it to be unaccepta bl e.
The fire barrier penetration system consists of passive devices _that must function in the event of a fire to mitigate the consequences.
There is no room for compromise on this issue. The licensee has agreed to accept the Technical Specifications as modified to remove the dependence on barrier size.
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