ML20246G207
| ML20246G207 | |
| Person / Time | |
|---|---|
| Site: | Limerick  |
| Issue date: | 08/25/1989 |
| From: | Murley T Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20246G199 | List: |
| References | |
| NUDOCS 8908310251 | |
| Download: ML20246G207 (76) | |
Text
7 g2. mt%
l l c' f
g NUCLEAR REGULATORY COMMISSION
.g r,
f cAssincraw. o. c. rosss Q..../
(
l PHILADELPHIA ELECTRIC COMPANY t
DOCKET NO. 50-353 LIMERICK GENERATING STATION, UNIT 2 FACILITY OPERATING LICENSE License No. NPF-85 1.
The Nuclear Regulatory Comission (the Comission or the NRC) has found l
that:
A.
The application for license filed by Philadelphia Electric Company (the lic66see) complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Comission's regulations set forth in 10 CFR Chapter I, and all required notifica-tions to other agencies or bodies have been duly made; B.
Construction of the Limerick Generating Station, Unit 2 (the facility) has been substantially completed in conformity with Construction Permit No. CPPR-107 and the application, as amended, the provisions of the Act and the regulations of the Comission; C.
The facility will operate in conformity with the application, as amended, the provisions of the Act, and the regulations of the Comission (except as exempted from compliance in Section 2.D. below);
D.
There is reasonable assurance:
(i) that the activities authorized by l
this operating license can.be conducted without endangering the health and safety of the public, and (ii) that such activities will be corducted in compliance with the Comission's regulations set I
forth in 10 CFR Chapter I (except as exempted from compliance in I
Section2.D.below);
E.
The licensee is technically qualified to engage in the activities authorized by this license in accordance with the Comission's regulations set forth in 10 CFR Chapter I; F.
The licensee has satisfied the applicable provisions of 10 CFR Part 4
140, " Financial Protection Requirements and Indemnity Agreemer.ts,"
l of the Comission's regulations; j
G.
The issuance of this license will not be inimical to the common defense and security or to the health and safety of the public; j
890S310251 690825 PDR ADOCK 05000353 P
FDC
A l "
r H. ' After weighing the environmental, ' economic, technical, and other benefits of the facility against environmental and other costs and e
considering available alternatives, the issuance of this Facility Operating License No. NPF-85, subject to the: conditions for protection of the environment set forth in the Environmental Protection. Plan attached as Appendix B,.is. in accordance with 20.CFR Part 51 of the Comission's regulations and all applicable requirements-have been satisfied; and I.
The receipt, possession, and use of source, byproduct and special nuclear material as authorized by this license will be in accordance.
with the Comission's regulations in 10.CFR Parts 30, 40 and 70.
2.
Based on.the, foregoing findings and the Decision of the Atomic' Safety and Licensing Board, LBP-85-25, dated July 22, 1985, the Comission's Order dated July 7,1989, and the Comission's Memorandum and Order dated August 25, 1989, regarding this facility, Facility Operating License NPF-85.is hereby issued to the Philadelphia Electric Company (the licensee), to read as follows:
.A.-
This license applies to the Limerick Generating Station, _ Unit' 2, a boiling water nuclear reactor and associated equipment, owned by
-Philadelphia Electric Company. The facility is located on'the licensee's site in Montgomery and Chester Counties, Pennsylvania on the banks of the Schuylkill River approximately 1.7 miles southeast of the city limits of Pottstown, Pennsylvania and 21 miles northwest of the city limits of Philadelphia, Pennsylvania, and is described in the ifcensee's Final Safety Analysis Report, as supplemented and amended, and in the licensee's Environmental Report-Operating License Stage, as supplemented and amended.
B.
Subject to the conditions and requirements incorporated herein, the Comission hereby licenses Philadelphia Electric Company:
(1) Pursuant to Section 103 of the Act and 10 CFR Part 50, to possess, use, and operate the facility at the designated location in Montgomery and Chester Counties, Pennsylvania, in accordance with the procedures and limitations set forth in this license; (2) Pursuant to the Act and 10 CFR Part 70, to receive, possess and to use at any time special nuclear material as reactor fuel, in accordance with the limitations for storage and amounts required for reactor operation, as described in the Final Safety Analysis Report, as supplemented and amended; (3) Pursuant to the Act and 10 CFR Parts 30, 40 and 70, to receive, possess and use at any time any byproduct, source and special nuclear material as sealed neutron sources for reactor startup, sealed sources for reactor instrumentation and radiation monitoring equipment calibration, and as fission detectors in amounts as required;
s
. e,
. (4) Pursuant to the Act and 10 CFR Parts 30 40 and 70, to receive, possess, and use in amounts as required any byproduct, source or special nuclear material without restriction to chemical or physical form, for sample analysis or _ instrument calibration or associated with radioactive apparatus'or components; and (5) Pursuant to the Act and 10 CFR Parts 30, 40 and 70, to possess, but not separate, such byproduct and special nuclear materials as may be produced by the operation of the faci ~1ty.
C.
This license shall be deemed to contain and is subject to the conditions specified in the Commission's regulations set forth in 10 CFR Chapter I (except as exempted f rom compliance in Section 2.D.
below) and is subject to all applicable provisions of the Act and to the rules, regulations, and orders of the Commission now or
~
hereafter in effect; and is subject to the additional conditions specified or incorporated below:
(1) ~ Maximum Power Level Philadelphia Electric Company is authorized to operate the facility at reactor core power levels of 3293 megawatts thermal (100 percent rated power) in accordance with the conditinns specified herein.
(2) Technical Specifications The Technical Specifications contained in Appendix A and the Environmental Protection Plan contained in Appendix B, are hereby incorporated into this license. PEco shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plcn.
(3) Fire Protection (Section 9,5, SSER 2)*
The licensee shall maintain in effect all provisions of the approved fire protection program as described in the Final Safety Analysis Report for the facility through Revision 58 and as approved in the SER through Supplement 9, and in the Fire Protection Evaluation Report through Revision 12, subject to the following provisions a and b below:
a.
The licensee shall make no change to features of the approved fire protection program which would decrease the level of fire protection in the plant without prior approval of the Commission. To make such a change the licensee must submit an application for license amendment pursuant to 10 CFR 50.90.
l l
l
- The parenthetical notation following the title of license conditions denotes the section of the Safety Evaluation Report and/or its supplements wherein the license condition is discussed.
L__________
?
- < b.
The licensee may make' changes to features of the approved
' fire protection program which.do not decrease the level of fire protection without prior' Comission approval after'-
such features have'been installed as approved, provided such changes do not otherwise involve a change in a. license condition or technical specification or result in an unreviewed safety question.(see'10 CFR 50.59).. However, the licensee.shall. maintain, in an auditable form, a current record of all.such changes including an~ evaluation of the effects of the change on the fire protection program and shall make such records available to NRC inspectors upon request. All. changes to the approved program made without prior Comission approval shall be reported to the Director of the 0?fice of Nuclear Reactor Regulation, together with supporting analyses, annually.
(4) Physica1' Security and Safeguards The licensee shall, fully implement and maintain in effect all provisions of the physical cecurity, guard training and qualification and safeguaro contingency plans previously approved by the Commission and all amendmcats and revisions to such plans made pursuant to the authority of.10 CFR 50.90 and 10 CFR 50.54(p). The plans, which contain Safeguards Information protected under 10 CFR 73.21, are entitled: " Limerick Generating Station, Units 1 & 2, Physical Security Plan," with revisions submitted through October 31,1988; " Limerick Generating Station, Units-1 & 2, Plant Security Personnel Training and Qualification Plan," with revisions submitted through October 1,1985; and
" Limerick Generating Station, Units 1 & 2. Safeguards Contingency Plan," with revisions submitted through November 15, 1986.
l D.
The facility requires exemptions from certain requirements of 10 CFR i
Part 50 and 10 CFR Part 70. These include (a) exemption from the i
requirementofparagraphIII.D.2.(b)(ii)ofAppendixJ,the testing of containment air locks at times when the containment 1
integrity is not required (Section 6.2.6.1 of the SER and SSER-3) 1 (b) exemption from the requirements of paragraphs II.H.4 and III.C.2 j
of Appendix J, the leak rate testing of the Main Steam Isolation
]
Valves (MSIVs) at the peak calculated containment pressure, Pa, and
{
exemption. from the-requirements of paragraph III.C.3 of Appendix J l
that the measured MSIV leak rates be included in the summation for the local leak rate test (Section 6.2.6.1 of SSER-3), (c) exemption l
from the requirement of paragraphs II.H.1 and III.C.2 of Appendix J,
)
1 1
l i
l
[
4
. the local leak rate testing of the Traversing Incore Probe Shear Valves (Section6.2.6.1oftheSERandSSER-3)[k)(1)rel (d) an exemption from the schedule requirements of 10 CFR 50.33 availability of funds for decommissioning the facility (Section 22.1, SSER 8) and (e) exemption from the requirement of 10 CFR Part 50.44, the inerting of containment six months after initial criticality (Section 6.2.5 of SSER 9). The special circumstances regarding exemptions (a), (b) and (c) are identified in Sections 6.2.6.1 of the SER and SSER 3.
An exemption from the criticality monitoring requirements of 10 CFR 70.24 was previously granted with NRC materials license No. SNM-1977 issued November 22, 1988. The licensee is hereby exempted from the requirements of 10 CFR 70.24 insofar as this requirement applies to the handling and storage of fuel assemblies held under this license.
These exemptions are authorized by law, will not present an undue risk to the public health and safety, and are consistent with the common defense and security. The exemptions in items a, b, c, d, and e above are granted pursuant to 10 CFR 50.12. With these exemptions, the facility will operate, to the extent authorized herein, in conformity with the application, as amended, the provisions of the Act, and the rules and regulatiou of the Commission.
E.
Except as otherwise provided in the Technical Specifications or Environmentti Protection Plan, the licensee shall report any violations of the requirements contained in Section 2.C of this license in the following manner:
initial notification shall be made within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> to the NRC Operations Center via the Emergency Notification System with written followup within thirty days in accordance with the procedures described in 10 CFR 50.73(b), (c),
and(e).
F.
The licensee shall have and maintain financial protection of such type and in such amounts as the Commission shall require in accordance with Section 170 of the Atomic Energy Act of 1954, as amended, to cover public liability claims.
S'!
.. G.
'This license is effective as of the date of issuance and shall-expire at midnight on. June 22, 2029.
FOR THE NUCLEAR REGULATORY COMMISSION
/S/
Thomas E. Murley, Director Office of Nuclear Reactor Regulation
Enclosures:
1.
Appendix A - Technical Specifications (NUREG-1376) 2.-
Appendix B - Environmental Protection Plan Date of. Issuance:
August 25, 1989
[ LIM /LIC.100]
PDI-2/P PDI-2/D NRR/PTSB NR gM-en RClark:
WButler Blambe ID Nash WEMas 89 of/o>/89 g /3/89 T /3 /E0 f/
89 7pB9 (hADRI[h
/
ADP h./
.co c g ERe BBoger gYat Jp/89 f/g/89
// _ glia JPart h APH for j;/
'0GC 8/16/89/..
8/D/89 89
@ da 47
/89 g/>
9
- i gi
.*f~'
+ :.,
y G.
.This license',is effective.as of the date of issuance and shall expire at midnight on-June'22, 2029.
.FOR THE NUCLEAR. REGULATORY COMMISSION-Thomas E. Murley, Director Office of Nuclear Reactor Regulation
Enclosures:
1.-
Appendix A - Technical Specifications (NUREG-1376) 2.
- Appendix'B
. Environment 1' Protection Plan
.Date of-Issuance:
[ LIM /LIC100]
PDI-2/LA PDI-2/PM PDI-2/D NRR/PTSB NRR/PTSB NRR/PTSB M0'Brien RClark:mr WButler Blambe IDinitz/DNash CThomas.
/. /89
/ /89
/ /89
/ /89
/ /89
/ /89 Ob ADRI DRPI/II ADP ADT Eje' BBoger SVarga JPartlow FMirag'sia
- T / g89
/ /89
/ /89
/ /89
/ /89 0NRR/DD ONRR/D JSniezek TMurley
/ /89
/ /89 1
l
a
.j.
.. G.
This license is effective as of the date of. issuance and shall expire at midnight on June 22, 2029.
FOR THE NUCLEAR REGULATORY COMMISSION 32-Thomas E. Murley, Director Office of Nuclear Reactor Regulation
Enclosures:
1.
Appendix A'- Technical Specifications (NUREG-1376) 2.
Appendix B - Environmental Protection Plan Date of Issuance: August 25, 1989 1
l l
!L_-____-_
4 NUREG-1376 Technical Specifications Limerick Generating Station, Unit No. 2 Docket No. 50-353 I
Appendix "A" to License No. NPF-85 Issued by the U.S. Nuclear Regulatory Commission Office of Nuclear Reactor Regulation August 1989
~
p ncoq e
p i,.
APPENDIX B TO FACILITY OPERATING LICENSE NO. NPF-85 LIMERICK GENERATIfl0 STATION UNITS 1 AND 2 PHILADELPHIA ELECTRIC COMPANY DOCKET NOS. 50-352 AND 50-353 ENVIRONMENTAL PROTECTION PLAN (NON-RADIOLOGICAL)
August 25, 1989
~.
hk, i
^
97
.. A
.1 ::
-LIMERICK GENERATING STATION UNITS:1 AND 2 ENVIRONMENTAL PROTECTION PLAN.
~
~
(NON-RADIOLOGICAL)
I TABLE OF CONTENTS Section Page 1.0 OBJECTIVES OF THE ENVIRONMENTAL PR9TECTION PLAN.....
1-1 1
l-l 2.0 ENVIRONMENTAL PROTECTION ISSUES.............
2-1 1
2.1 Aquatic Issues......................
2 L 2.2 Terrestrial Issues....................
2-2 2.3 No i se I s sues........................
2-2 3.0 CD itSTENCY REQUIREMENTS.................
3-1 3.1 Plant. Design and'0peration................
3-1 l'
3.2 Reporting Related to the NPDES Permit and State Certifications...................... 3-2 3.3 Changes Required for Compliance with Other Environmental Regulations...............
3-2 4.0 ENVIRONMENTAL CONDITIONS.................
4-1 4.1 Unusual or Important Environmental Events........
4-1 4.2 Environmental Monitoring.................
4-1 5.0 ADMINISTRATIVE PROCEDURES................
5-1 5.1 Review and Audit.....................
5-1 5.2 Records Retention....................
5-1 5.3 Changes in Environmental Protection Plan.........
5-1 5.4 Plant Reporting Requirements...............
5-2
,9 I
1.0 OBJECTIVES OF THE ENVIRONMENTAL PROTECTION PLAN The Environmental Protection Plan (EPP) is to provide for protection of
~
non-radiological environmental values during operation of the nuclear facility.
The principal objectives of the EPP are as follows:
(1) Verify that.the facility is operated in an environmentally acceptable manner, as established by the Final Environmental Statement-Operating License Stage (FES-OL) and other NRC environmental impact assessments.
(2) Coordinate NRC requirements and maintain' consistency with other Federal, State and local requirements for environmental protection.
(3) Keep.NRC informed of the environmental effects of facility construction and operation and of actions taken to control those effects.
. Environmental concerns identified in the FES-OL which relate to water quality matters are regulated by way of the licensee's NPDES permit.
l 1-1 l
t
?-
4
?.
2.0': ENVIRONMENTAL PROTECTION ISSUES
'In the:FES-OL dated April, 1984, the staff. considered the environmental impacts associated'with the operation of theltwo unit Limerick Generating l'
' Station. Certain environmental issues were identified which required study or'.
' license conditions to resolve environmental concerns and to assure adequate protection of the environment.
l
- 2.1 Aquatic Issues (1) During operation, the station blowdown temperature will exceed the maximum permissible
- temperatures ~ set by the applicable water quality standards.
However, the affected area of the Schuylkill P,iver is expected to be' smaller than the maximum area permitted by the Delaware River Basin Commission. -(FESSection5.3.2.2)
(2) The water quality of the station discharge, after initial mixing with the Schuylkill River, is predicted to, at times, exceed the applicable quality criteria for some constituents, based on source water maximum constituent' concentrations. These exceedances are expected for constituents whose maximum river concentrations also exceed the applicable criteria.
(FES Section 5.3.2.3)
(3) Chlorination of station cooling waters for condenser and cooling tower biofouling control may result in some adverse impacts to Schuylkill River biota in the vicinity of the station discharge.
(FESSection5.3.2.3)'
(4) Operation of the Point Pleasant Diversion will alter the hydrology, aquatic habitats, and water quality of the headwater section of the East Branch of Perkiomen Creek but the diversion waters are expected to provide beneficial dilution of waste loads enteriag the stream in its middle and lower reaches.
(FES Sections 5.3.2.3 and 5.2.2)
(5) The supplemental cooling water withdrawal from Perkiomen Creek using i
state-of-the-art technology will result in localized effects from entrainment of fish larvae.
(FES Section 5.5.2) 2-1
[
Q_;
h 2.2 Terrestrial Issues-No specific terrestrial issues were identified by the NRC staff in the FES-OL.
2.3 Noise Issues-p-
(1)- Tones from the Point Pleasant pumphouse transformers are predicted to be audible and may cause annoyance at a nearby residence. Noise. monitoring and, if necessary, mitigative measures to make the tones inaudible have been mandated by the ASLB.
(FES Sections 5.12.1 and 5.14.4.1)
(2) Noise from transformers:and pumps in the Bradshaw Reservoir pumphouse may be audible at nearby residences. The licensee has committed to ambient-and operational noise level monitoring and implementation of identified mitigative measures, if necessary, to reduce noise levels below those likely to cause annoyance and complaints.
(FES Sections 5.12.2 and 5.14.4.2)
(3) Offsite noise. levels in the vicinity of the Limerick site during station operation are not expected to be high enough above ambient levels to annoy nearby residents. But because of uncertainties in the assessment, a confirmatory noise monitoring program and implementation of mitigative measures, if necessary, will be undertaken.
(FES Sections 5.12.3 and 5.14.4.3)
NRC requirements with regard to noise issues are specified in Section 4.3 of this EPP.
2-2 i
9
- i 3.0 CONSISTENCY. REQUIREMENTS-3.1 Plant Design and Operatioit
' The licensee may make changes in station design or operation or perform tests or experiments affecting the environment provided such activities do not involve an unreviewed environmental question and do not involve a change in the EPP.* Changes in station design or operation or performance of tests or experiments which do not affect the environment are not subject to the requirements of this EPP. Activities governed by Section 3.3 are not subject-to the requirements of this Section.
Before engaging.in additional construction or operational activities which may significantly affect the environment, the licensee shall prepare and record an environmental evaluation of such activity. Activities are excluded from this requirement if all measurable non-radiological environmental effects are confined to the on-site areas previously disturbed during site preparation and l
plant construction. When the evaluation indicates that such activity involves an unreviewed environmental question, the licensee shall provide a written evaluation of such activity and obtain prior NRC approval. When such activity involves a change in the EPP, such activity and change to the EPP may be implemented only in accordance with an appropr.iate license amendment as set forth in Section 5.3 of this EPP.
A proposed change, test or experiment shall be deemed to involve an unreviewed i
environmental question if it concerns:
(1) a matter which may result in a significant increase in any adverse environmental impact previously evaluated in the FES-OL, environmental impact appraisals, or in any decisions of the Atomic Safety and Licensing Board; or (2) a significant change in effluents or l
1 l
- This provision does not relieve-the licensee of the requirements of 10 CFR 50.59 3-1
d f
1 power-level; or (3) a matter not previously reviewed and evaluated in-the documents specified in (1) of this Subsection, which may have a significant j
~
adverse environmental' impact.
The licensee shall maintain records of changes in facility design or operation l
and of tests and experiments carried out pursuant to this Subsection. These records shall-include written evaluations which provide bases for the determination that the change, test, or experiment does not involve an unreviewed environmental question or constitute a decrehse in the effectiveness of the EPP to meet the objectives specified in Section 1.0.
The licensee shall include as part of the Annual Environmental Operating Report (per Subsection 5.4.1) brief descriptions, analyses, interpretations and evaluations of such changes, tests and experiments.
3.2 Reporting Related to the NPDES Parmit and State Certification Changes to, or renewals of, the NPDES Permit or the State certification shall be reported to the NRC within 30 days following the date the change or renewal is approved.
If a permit or certification, in part or in its entirety, is appealed and stayed, the NRC shall be notified within 30 days following the date the stay is granted.
The licensee shall notify the NRC of changes to the effective NPDES Permit proposed by the licensee by providing the NRC with a copy of the proposed change at the same time it is submitted to the permitting agency. The licensee shall provide the NRC a copy of the application for renewal of the NPDES Permit at the same time the application is submitted to the permitting agency.
3.3 Changes Recuired for Compliance with Other Environmental Regulations Changes in plant design or operation and performance of tests or experiments which are required to achieve compliance with other Federal, State and local environmental regulations are not subject to the requirements of Section 3.1.
3-2
l 1%
4 ~. 0 ENVIRONMENTAL CONDITIONS.
4.1 linusual. or Important Environmental Events Any occurrence of an unusual or important event that indicates or could result in significant environmental impact causally related to plant operation shall be recorded and reported to the NRC within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> followed by a written report per Subsection 5.4.2.
The following are examples: excessive
' bird impaction events, onsite plant or animal disease outbreaks, mortality or unusual occurrence of any species protected by the Endangered Species Act of 1973, fish kills,. increase in nuisance organisms or conditions, and unanticipated or einergency discharge of waste water or chemical substances.
No routine monitoring programs are required to implement this condition.
4.2 Environmental Monitoring 4.2.1 Aquatic Monitoring The certifications and permits required under the Clean Water Act provide mechanisms for protecting water cuality and, indirectly, aquatic biota. The NRC will rely on the decisions made by the Conanonwealth of Pennsylvania, under the authority of the Clean Water Act, for any requirements for aquatic monitoring.
4.2.2 Terrestrial Monitoring No terrestrial monitoring is required.
1 4.2.3 Maintenance of Transmission Line Corridors 1
l-The use of herbicides within the Limerick Generating Station transmission line corridors (Limerick to Cromby, Cromby to Plymouth Meeting, Cromby to
{
l-North Wales, and Limerick tc Whitpain) shall conform to the approved use of selected herbicides as registered by the Environmental Protection Agency and
-approved by Commonwealth authorities and applied as directed on the pesticide
- label, j
4-1 1
)
l Records shall be maintained in the licensee's Electric Transmission and Distribution Department concerning herbicide use. Such records shall include the following information:
commercial and chemical names of materials used; concentration of active material in formulations for field use; diluting substances other than water; rates of application; method of application; location; and the date of application. Such records shall be maintained for a period of 5 years from the date they are prepared and shall be made readily available to'the NRC upon request. There shall be no routine reporting requir? ment associated with this condition.
4.2.4 Noise Monitoring 4.2.4.1 General Requirements for Sound Level Surveys Surveys shall be conducted to quantify the ambient (i.e., background) and the operational sound levels that exists at various locations in the vicinity of the LGS site and supplemental cooling water system facilities, as specified below. The ambient sound level surveys shall be conducted during the time period when significant station or supplemental cooling water system outdoor construction activity has ended, and with no operation of the cooling system or other significant noise sources (e.g., transformers) so that measured sound levels are not affected by such activities.
The operational sound level suryeys shall be conducted as soon as practicable within the first year of the operational phase of the various station associated facilities, when transformers are energized or when the cooling towers are operating with their design water flow rates, as appropriate for the facility under study.
For the LGS site, one survey shall be conducted for one unit normal operation and a second survey for two unit normal operation.
The conduct of the surveys for both phases shall be such that the results are comparable.
For each of the surveys, except as otherwise specified below or in LBP-83-11, sound level data shall be collected at several sites, the exact number and location to be selected by the licensee after consideration of (1) existing onsite and nearby offsite noise sources and barriers; (2) noise sensitive land uses in the site vicinity (e.g., residences, schools, churches, cemeteries, hospitalt, parks); and (3) previously conducted noise surveys in the site vicinity.
4-2
K.; +
V,,
4 The selection, calibration and use of equipment, conduct of the surveys, and the analysis and reporting'of data shall conform to the provisions of the
- applicable American Nationa.1. Standards Institute Standards.
The results' of the surveys conducted under this program shall be summarized, interpreted and reported in accordance with Section 5.4.1 of.this EPP.
The final report of this program shall present a brief assessment by the
' licensee of. the environmental impact and supplemental cooling water system operation on the various offsite acoustic environments, and shall describe the mitigative measures, if any, that have been, or are to be taken to reduce the impact of plant or supplemental cooling water system noise levels on the
.offsite environments. This report shall also contain a list of noise-related complaints or inquiries received by Philadelphia Electric Company concerning the. Limerick Generating Station or its supplemental cooling water system
. subsequent.to issuance of the operating license along with a description of the action taken by Philadelphia Electric Company to resolve these complaints or inquiries.
f This program shall terminate upon completion of the collet. tion of.the specified sound level data for each phase and submission of an acceptable final report.
4.2.4.2 Point Pleasant Pumphouse An ASLB ruling (LBP-83-11; March 8,1983) requires that the licensee conduct a one-time field study after the transformers are placed in operation at Point Pleasant. The noise from operation of the transformers shall be reduced to a level so that the transformer core tones will be inaudible (i.e., not above the masking level, as defined below) at the site boundary.
The licensee shall determine, based on onsite measurements o the delta L(ex)
(i.e., the noise level in excess of the masking level) for each tone. The masking level is defined as "N" dB above the ambient spectrum level, where "N" is defined as follows:
4-3 l
n.
h
' i g. 5 4,ca p ~ ' r r
MaskingLevel"N"L Tone Frequency, dB' Hz 17.5 120 17.4 240 17.8 360 18.2 480 The.stsps't' be carrlied.out are:
o
~
, (1) Measures'thebroadband-(nontonal)ambientnighttime(midnightto4'a.m.)
levels in all octave bands of frequency when the transformers.are not energized (step A in the ASLB ruling). From these, tone-masking sound levels (thresholds of audibility) at each tone frequency shall be calculated.
- (2).Measurethelevelofthet'ones.thatexistwhenthetransformersare energized (stepB)intheone-third-octavebandsthatcontainthemfor comparison with the calculated masking levels from step A to determine audibility. Transformer tones should not be greater than the masking
- level for specific one-third-octave bands If, for any reason, it is not practicable to de-energize the transformers at any
' time during'the prescribed measurement time period (midnight to 4 a.m.),
alternative measurements shall be made.
In that event, only one measurement y
of the entire spectrum need be made, but it must be done entirely in one-third-octave bands. Any measured octave band sound level that includes a tone or. tones cannot be used to determine the true background ambient sound level for that octave band.
In such cases, the background ambient level in any one-third-octave band containing a tone shall be approximated by interpolating between those adjacent one-third-octave band levels that do not contain any tones.
If audible tones, defined as any delta L(ex) value greater than 0, are found at the site boundary, mitigative measures, such as a three-sided barrier or a 4-4 i
h m_m____.___.___E______.-_______________m.
l N
full enclosure around the transformers, shall be implemented (and their performance verified by noise measurement) so that tones are not greater than the masking level as previously determined.
4.2.4.3 Bradshaw Reservoir The noise measurements proposed for the Point Pleasant site shall be made at the Bradshaw pumphouse site boundary while energized, on the line between the transformers and residences B, D, and F, to determine if the tones would be audible at those points (see Figure 5.6, LGS Final Environmental Statement, NUREG-0974,p.5-140). Measures to render these tones inaudible to these points shall then be applied, as necessary, and their performance verified.
The measurement program recommended for Bradshaw to test for audible tones from the transformers shall also be extended to include audible tones in the 1000-Hz octave band at the southeast site boundary in the direction of residence B.
4.2.4.4 Limerick Site The operational measurement program to be applied to the Point Pleasant area shall be applied to the area immediately beyond the northern boundary of the Limerick site as well.
If audible tones are found to be present at the northern site boundary after Unit I transformers are energized, mitigative measures shall be taken and their p'Erformance verified, as required for the Bradshaw Reservoir, to cause those tones to be inaudible.
For the LGS site only, data shall be collected from each sampling site during the time of year when foliage of deciduous trees is present and also during the time of year when such foliage is largely absent. Data collected from each sampling location shall encompass both the daytime and nighttime periods.
i i
4-5 3
5.0 ADMINISTRATIVE PROCEDURES 5.1 Review and Audit Ti.e licensee shall provide for review and audit of compliance with the EPP.
The audits shall be conducted independently of the. individual or groups responsible for performing the specific activity. A description of the organizational structure utilized to achieve the independent review and audit function and results of the audit activities shall be maintained and made available for inspection, 5.2 Records Retention Records and logs relative to the environmental aspects of station operation shall be made and retained in a manner convenient for review and inspection.
These records and logs shall be made available to NRC on request.
Records of modifications to station structures, systems and components determined to potentially affect the continued protection of the environment shall be retained for the life of the station. All other records, data and logs relating to this EPP shall be retained for five years or, where applicable, in accordance with the requirements of other agencies.
5.3 Changes in Environniental Protection Plan Request for changes in the EPP shall include an assessment.of the environmental impact of the proposed change and a supporting justification.
Implementation of such changes in the EPP shall not commence prior to NRC approval of the proposed changes in the form of a license amendment incorporating the appropriate revision to the EPP.
i 5-1 i
Y 5.4 Plant Reporting Requirements 5.4.1
. Routine Reports An Annual Environmental Operating Report describing implementation of this EPP for the previous year shall be submitted to the NRC prior to May 1 of each l
year. The initial report shall be submitted prior to May 1 of the year following issuance of the operating license.
The report shall include summaries and analyses of the results of the environ-l-
' mental protection activities as required by Subsection 4.2 of this EPP for the report period, including a comparison with related preoperational studies, operational controls (as appropriate), and previous non-radiological environmental monitoring reports, and an assessment of the observed impacts of the plant operation on the environment.
If harmful effects or evidence of trends towards irreversible damage to the environment are observed, the licensee shall provide a detailed analysis of the data and a proposed course of.
mitigating action.
The Annual Environmental Operating Report shall also include:
(1) A list of EPP noncompliance and the corrective actions taken to remedy them.
(2) A list of all changes in station design or operation, tests, and experiments made in accordance with Subsection 3.1 which involved a potentially significant unreviewed environmental question.
(3) A list of nonroutine reports submitted in accordance with Subsection 5.4.2.
In the event that some results are not available by the report due date, the report shall be submitted noting and explaining the missing results. The missing results shall be submitted as soon as possible in a supplementary report.
5-2
- s
,..f '
5.4.2
-Nonroutine Reports A written' report shall-be subadtted to the NRC within 30 days of occurrence of a nonroutine event. The report shall (a) describe, analyze, and evaluate the event, including extent and magnitude of the impact and plant operating characteristics, (b) describe the probable cause of the event, (c) indicate the action taken to correct the reported event, (d) indicate the corrective action taken to preclude repetition of the event and to prevent similar occurrences _
involving similar components or systems, and (e) indicate the agencies notified and their preliminary responses.
Events reportable under this subsection which also require reports to other Federal, State'or local agencies shall be reported in accordance with those reporting requirements in lieu of the requirements of this subsection. The NRC shall be provided a copy of such' report at the same time it is submitted to the other agency.
l l
5-3
1 (w }'
O.
7590-01 UNITED STATES NUCLEAR REGULATORY COMMISSION DOCKET NO. 50-353 LIMERICK GENERATING STATION; UNIT NO. 2 NOTICE OF ISSUANCE'0F FACILITY OPERATING LICENSE Notice is hereby given that the U. S. Nuclear Regulatory Comission (the Comission), has issued Facility Operatir:g. License No.. NPF-85 to the
- Philadelphia Electric Company.-(the licensee) which authorizes operation of
. the Limerick Generating Station, Unit No. 2 (the facility), by Philadelphia Electric Company at reactor core power levels of 3293 megawatts thermal in-accordance with the provisions of the License, the Technical Specifications and the Environmental Protection Plan.
The Limerick Cenerating Station, Unit No. 2, is a boiling water nuclear reactor located on the licensee's site in Montgomery and Chester Counties, Pennsylvania on the banks of the Schuylkill River approximately 1.7 miles southeast of the city limits of Pottstown, Pennsylvania and 21 miles' northwest of the city limits of Philadelphia, Pennsylvania.
The application for the license complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act) and the Comission's regulations.- The Comission has made appropriate findings as required by the Act and the Commission's regulations in 10 CFR Chapter I, which are set forth in'the License.
Prior public notice of the overall action involving the j
proposed issuance of an operating license was published in the Federal Register I
on August 21, 1981 (46 FR 42557 - 42558).
1
P 1
.a
'.i n;
L'.. The Comission has determined that the fosuance of this license will not result in any environmental impacts other than those evaluated in the Final
~
Environmental Statement since ine activity authorized by the license is encompassed by the overall action evaluated in the Final Environmental Statement.
Pursuant to 10 CFR 51.32, the Comission has determined that the issuance of the exemptions included in this license will have no significant impact on the environment (54 FR 15851), (54 FR 24607) and (54 FR 33298).
For Further details in respect to this action, see (1) Facility Operating License NPF-85' complete with Technical Specifications and the Environmental Protection P1an;-(2) the final report of the Advisory Committee on Reactor Safeguards, dated May 11, 1989; (3) the Comission's Safety Evaluation Report, I
dated August 1983 (NUREG-0991), Supplements 1 through 9; (4) the Final Safety Analysis Report and Amendments thereto; (5) the Environmental Report and 1
l supplements thereto; (6) the Final Environmental Statement dated April 1984 (NUREG-0974); (7) the Atomic Safety and Licensing Board Decision, LBP-85-25, dated July 22, 1985; (8)- the Comission's Order dated July 7,1989, and (9) the Commission's Memorandum and Order dated August 25, 1989.
These items are available for public inspection at the Commission's Public
~
Document Room, 2120 L Street, N.W., Washington, D. C.
20555, and at the l
Pottstown Public Library, 500 High Street, Pottstown, Pennsylvania, 19464 l
A copy of Facility Operating License NPF-85 may be obtained upon request I
addressed to the U. S. Nuclear Regulatory Comission, Washington, D. C.
- 20555, Attention: Director, Division of Reactor Projects I/II. Copies of the Safety Evaluation Report and its Supplements 1 through 9 (NUREG-0991) and the Final Environmental Statement (NUREG-0974) may be purchased through the U.S. Government
1
,..,f. t '
3 Printing.0ffice by calling (202) 275-2060 or by writing to the U.S. Government l
Printing Office, P. O. Box 37082, Washington, D.C.
20013-7082. Copies may also i
be purchased from the National Technical Information Service, U. S. Department of Commerce, 5285 Port Royal Road, Springfield, Virginia 22161.
Dated at Rockville, Maryland, this 25th day of August 1989.
FOR THE NUCLEAR REGULATORY COMMISSION Walter R. Butler, Director Project Directorate I-2 Division of Reactor Projects I/II Office of Nuclear Reactor Regulation l
l
/
UNITED STATES
[
g NUCLEAR REGULATORY COMMISSION k.....,[j g
wAswiwaroN, D. C. 20666 August 25, 1989 Docket Nos. 50-352, 50-353 Amendment to Indemnity Agreement No. B-101 Amendment No. 6 Effective 8/25/89, Indemnity Agreement No. B-101, between Philadelphia Electric Company and the Nuclear Regulatory Conunission, dated April 3,1984, as amended, is hereby further amended as follows:
Item 3 of the Attachment to the indemnity agreement is deleted in its entirety and the following st,bstituted therefor:
Item 3 -- License number or numbers SNM-1926 (From 12:01 a.m., April 3, 1984, to 12 midnis t, October 25, 1984 inclusive NPF-27 (From 12:01 a.m., October 26, 1984, to 12 midnig t, August 7, 1985 inclusive NPF-39 (From12:01a.m., August 8,1985)
SNM-1977 (From 12:01 a.m., November 22, 1988, to 12 midnig t, June 21, 1989 inclusive NPF-83 (From 12:01 a.m., June 22, 1989, to 12 midnig t, July 9, 1989 inclusive NPF-84 (From 12:01 a.m., July 10, 1989, to 12 midnig t, August 24, 1989 inclusive NPF-85 (From 12:01 a.m.,
August 25, 1989
)
l i
1
{
l i
l
n,_-_.
1 2
FOR THE UNITED STATES NUCLEAR REGULATORY COMMISS10tl
{s,'. M, w, Y --
Cecil 0. Thomas, Chief Policy Development and Technical Support Branch Program Management, Policy Development and Analysis Staff Office of Nuclear Reactor Regulation Accepted 1989 By Philacic1phia Electric Company 4
O i
l l
)
l l
E______.______
i
7 NUREG-0991 Supplement No. 9 Safety Evaluation Report related to the operation of Limerick Generating Station, Units 1 and 2 Docket Nos. 50-352 and 50-353 Philadelphia Electric Company 1
~
U.S. Nuclear Regulatory Commission Office of Nuclear Reactor Regulation August 1989
,gaascy ikf}
w
(_
O NUREG-0991 i.
Supplement No. 9 Safety Evaluation Report re.atec to the operation of Limerick Generating S':ation, Units :L and 2 Docket Nos. 50-352 and 50-353 Philadelphia Electric Company U.S. Nuclear Regulatory Commission Omce of Nuclear Reactor Regulation August 1989
(
g g
P' NUREG-0991 l
' Supplement No. 9 Safety Evaluation Report L
related to the operation of l
Limerick Generating Station,
~ Units 1 and 2
~
Docket Nos. 50-352 and 50-353 Philadelphia Electric Company l
l U.S. Nuclear Regulatory Commission Office of Nuclear Reactor Regulation August 1989 p aeoq N!51
09 ABSTRACT In August 1983 the staff of the Nuclear Regulatory Commission issued its Safety Evaluation Report (NUREG-0991) regarding the application of the Philadelphia Electric Company (the licensee) for licenses to operate the.
Limerick Generating Station, Units I and 2, located on a site in Montgomery and Chester Counties, Pennsylvania.
Supplement I to NUREG-0991 was issued in December 1983. Supplements 2 and 3 were issued ir October 1984. License NPF-27 for the low-power operation of Limerick Unit I was issued on October 26, 1984. Supplement 4 was issued in May 1985, Supplement 5 was. issued in July 1985, and Supplement 6 was issued in August 1985. These supplements addressed.further issues that required resolution before Unit 1 proceeded beyond the 5-percent p)ower level.
The full-power operating license for Limerick Unit 1 (NPF-39 was issued August 8, i
1985, and the unit has completed two cycles of operation.
Supplement 7 was issued April 1989 to address some of the few significant design differences between Units I and 2, the resolution of issues that remained open when the Unit I full-power license was issued and an assessment of some of the issues that required resolution *before issuance of an operating license for Unit 2.
Supplement 8, issued in June 1989 resolved all the issues necessary to support the issuance of a low power license for Unit 2.
Operating license NPF-83, authorizing Unit 2 to load fuel and conduct pre-criticality testing, was issued on June 22, 1989. Operating license NPF-84, authorizing centinued testing and operation of Limerick Unit 2 at power levels up te five percent (5%),wasissuedonJuly 10, 1989.
This document, the ninth supplement to the SER (SSER-9), also primarily relates to Unit 2.
This supplement addresses the remaining issues that required resolution before issuance of a full power ifcense for Unit 2.
Limerick SSER 9 iii
L.;
V i
w q:
TABLE OF CONTENTS l
3a,le.
1' L
ABSTRACT.............................................................
111 1-INTRODUCTION AND GENERAL DESCRIPTION OF PLANT...................
1-1 1.1 Introduction............................................... 1-1 3
DESIGN CRITERIA FOR STRUCTURES, SYSTEMS, AND COMPONENTS......... 3-1 3.9 Mechanica l Systems and Components.......................... 3-1 3.9.3 ASME Code Class 1, 2 and 3 Components, Component Supports, and Core Support Structures.......... 3 3.9.3.5 Bulletin 88-05.........................
3-1 3.10 Seismic and Dynamic Qualification of Mechanical and Electrical Equipment................................... 3-1 3.11 Environmental Qualification of Electrical Equipment Important to Safety and Safety-Related Mechanical Equipment.................................................. 3-2 3.11.1 Introduction.....................................
Background....................................... 3-2 3.11.2 3 3.11.2.1 Purpose................................ 3-3 3.11.2.2 Scope.................................. 3-3 3.11.3 Staff Evaluation.................................
3-3 3.11.4 Conclusions,..................................... 3-4 4
REACT 0R......................................................... 4-1 4.4 Thermal-Hydraulic Design................................... 4-1 4.4.4 The rma l-Hydrau l ic Stab 11i ty...................... 4-1 d'
5 REACTOR COOLANT SYSTEMS......................................... 5-1 5.3 Rea c t o r Ve s se l............................................. 5 -1 5.3.1 Reactor Yessel Materials......................... 5-1 5.3.1.2 Fracture Toughness..................... 5-1 5.3.2 Pressure-Temperature Limits...................... 5-2 6
ENGINEERED SAFETY FEATURES................................
6-1 6.2 C o n t a i nme n t Sy s t em s........................................ 6 -1 6.2.5 Combustible Gas Contro1.......................... 6-1 1
7 INSTRUMENTATION AND CONTR0LS....................................
7-1 7.4 Systems Required for Safe Shu tdown......................... 7-1 7.4.2 Specific Findings....................................
7-1 7.4.2.1 Capability for Safe Shutdown Following Loss of Electrical Power to Instrumentation and Controls............. 7-1 7.4.2.1.1 Common Power Source Failure A n a ly s i s.......................... 7 -2 7.4.2.1.2 Common Sensor Failure Analysis.... 7-2 7.4.2.1.3 HELB and Affected NSCS Component......................... 7-2 7.4.2.1.4 Conclusions....................... 7-3 Limerick SSER 9
-v-i
n:
i l
~
l 1
TABLE OF CONTENTS (Continued) 7.7 Control Systems............................................. 7-3
{
.7.7.2 Specific Findings................................. 7-3 7.7.2.1 High-Energy-Line-Break and Consequential
'J s
Control Systems Failures (IE Information Notice 79-22) and Multi Systems Failures.......ple Control
............... 7-3 15 ACCIDENT ANALYSES............................................... 15-1 15.8 An ti cipa ted Transient Wi thout Scram........................ 15-1 15.8.1 Introduction..................................... 15-1 15.8.2 Review Criteria..................................
15-2 15.8.3 ARI and RPT System Description................... 15-2 15.8.4 Eva lua tion of ARI Sy s tem......................... 15-3 15.8.4.1 Safety Related Requirements (IEEE Standard 279)..........................
15-3 15.8.4.2 Redundancy............................. 15-4 15.8.4.3 D i vers i ty f rom Ex i s t i ng RTS.......,.... 15-4 15.8.4.4 Physical Separation from Existing RTS........................... 15-4 15.8.4.5 Environmental Qualification............ 15-4 15.8.4.6 Seismic Qualification.................. 15-5 15.8.4.7 Quali ty As su ra n ce...................... 15-5 15.8.4.8 Safety Related (IE) Power Supply....... 15-5 15.8.4.9 Testability at Power...................
15-5 15.8.4.10 Inadvertent Actuation.................. 15-5 15.8.4.11-Manual Initiation...................... 15-5 15.8.4.12 Information Readout....................
15-6 15.8.4.13 Completion of Protective Action
%s Once Initiated.........................
15-6 15.8.4.14 Maintenance Bypass..................... 15-6 15.8.4.15 Conclusion.............................
15 15.8.5 Eva luation of ATWS/RPT Sys tem.................... 15-6 15.8.5.1 Safety Rela ted Requi rements............ 15-6 15.8.5.2 Redundancy............................ 15-6 15.8.5.3 Diversity from Ex isting RTS........... 15-6 15.8.5.4 Physical Separation from Existing RTS...
..................... 15-7 15.8.5.5 Environmental Qua lification........... 15-7 15.8.5.6 Seismic Qualification.................. 15-7 15.8.5.7 Quality Assurance......................
15-7 15.8.5.8 Safety Related (IE) Power Supply.......15-7 15.8.5.9 Testability at Power.................. 15-7 15.8.5.10 Inadvertent Actuation..................
15-7 15.8.5.11 Conclusion on ATWS RPT System..........15-8 15.8.6 Evaluation of SLCS............................... 15-8 15.8.6.1 Safety Related Requirements............ 15-8 15.8.6.2 Evaluation.............................
15-8 15.8.6.3 Conclusion on SLCS..................... 15-8 15.8.7 Technical Specifications......................... 15-9 15.8.8 Conclusions......................................
15-9 Limerick SSER 9
-vi-
p.
u TABLEOFCONTENTS(Continued) 1 p
16 TECHNICAL SPECIFICAT0NS......................................... 16-1 n
17' QUALITY ASSURANCE............................................... 17-1 17.6. Readiness Verifica tion Program............................ 17mi -
17.6.1 Independent Construction Assessment (ICA)..........17-1 1
17.6.2 Independent Design Assessment (IDA)................ 17-4 17.6.3 Conclusions........................................ 17-5 APPENDICES A
CHRONOLOGY H
PRINCIPAL STAFF CONTRIBUTORS U
ERRATA TO THE SAFETY EVALUATION REPORT FOR THE LIMERICK GENERATING STATION Limerick SSER 9
-vii-l t
-1 INTRODUCTION AND GENERAL DESCRIPTION OF PLANT 1.1 Introduction In August-1983, the Nuclear Regulatory Commission (hereinafter referred to as the NRC or the staff) issued its Safety Evaluation Report (SER) NUREG-0991, regardingtheapplicationofthePhiladelphiaElectricCompany[ hereinafter referred to as PECo or the licensee) for lic&nses to operate the Limerick Generating Station, Units 1 and 2, Docket Nos. 50-352 and 50-353. Supplement I to the SER was issued in December 1983, Supplements 2 and 3 were' issued in October 1984, and Operating License NPF-27, authorizing power.up to 5 percent, was issued on October 26, 1984. Supplement 4 to the SER was issued in May 1985, Supplement 5 was issued in July 1985, and Supplement 6 was issued in August 1988. These supplements addressed issues that required further resolution before Unit 1 proceeded beyond the 5-percent power level. A full-power operating license (NPF-39) was issued for Limerick Unit 1 on-August 8, 1985.
As noted above, the staff's SER assessed operation of both Limerick Units 1 and 2.
Construction of Unit 2 was halted in January 1984 by Order of the Pennsylvania Public Utility Commission. At the time, construction was_about 30 percent complete. Construction of Unit 2 resumed in February 1986 with PEco's agreement to accept a cost containment cap of about $3.1 billion for construction and certain operational incentive programs. On May 3, 1988, the Commission modified Construction Permit CPPR-107 to extend the earliest and latest. completion dates to May 1,1989, and January 1,1992, respectively.
Supplement 7 to the SER was issued April 1989 and primarily related to Unit 2.
SSER-7 addressed 'some of the few significant design differences between Units 1 and 2, the resoluticn of issues that remained open when the Unit I full-power license was issued and an assessment of some of the issues that required resolu-tion before issuance of a low-power operating license for Unit 2.
Supplement 8 to the SER was issued June 1989 and also addressed primarily Unit 2 issues that
.5' required resolution before issuance of a low power license. Operating License NPF-83, authorizing Unit 2 to load fuel and conduct pre-criticality testing, was issued on June 22, 1989. Operating license NPF-84, authorizing continued testing and operation for Limerick Unit 2 at power levels up to five percent (5%), was issued on July 10, 1989.
This document, the ninth supplement to the SER (SSER-9), also primarily relates to Unit 2.
This supplement addresses the remaining issues that require resolu-tion before issuance of a full power operating license for Unit 2.
I l
L Limerick SSER 9 1-1 u-___---__--_--___
Each of the sections and appendices of this supplement is-numbered the same as the'related portion of the SER. Each section complements the discussion in the SER and Supplements ' through 8, unless otherwise noted. Appendix A is a continuation of the chronology of this safety. review. Appendix H lists the principal contributors. Appendix U is an errata which provides minor corrections of SSER-8.
Copies of this supplement are available for inspection at the NRC Public Document Room, 2120 L Street, N.W., Washington, D.C. and at the local Public Document Room at the Pottstown Public Library, 500 High Street, Pottstown, Pennsylvania 19464.
The NRC Project Manager for Limerick Units 1 and 2 is Richard J. Clark. He may be contacted by telephone at (301) 492-3041 or by mail at the following address:
Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Washington, DC 20555
)
i Limerick SSER 9 1-2
1 3 DESIGN CRITERIA FOR STRUCTURES, SYSTEMS, AND COMPONENTS 3.9 Mechanical Systems and Components 3.9.3 ASME Code Class 1, 2 and 3 Components, Component Supports, and Core Support Structures j
3.9.3.5 Bulletin 88-05 In the safety evaluation transmitted on June 20, 1986 and in SSER 8, the staff provided the results of our review of the licensee's submittals of March 31 and June 2,1989 for Limerick 2.
In each of these, we indicated that the licensee had conducted-stress analysis on 52 installed safety-related items which were found to have tensile strengths below 66 ksi (396 L converted to 137 BHN). These52itemswereidentifiedas46carbonsteelf9angesand6 stainless steel flanges. Actually, stress analysis was only conducted on the 46 carbon steel flanges. The 6 stainless steel flanges were found to be acceptable based on compliance with the criteria in Section 5 of NUMARC's Generic Testing Program Response of October 1988, as previously indicated in section 3.3 of our June.20, 1989 safety evaluation.
3.10 Seismic and Dynamic Qualification of Mechanical and Electrical Equipment By letter dated March 7,1989, the licensee submitted a report to provide confirmation that the Equipment Qualification Programs used for dynamic and environmental qualification of Limerick Generating Station Unit 2 are consistent with the programs used for Unit 1.
While the programs are consistent, some equipriient items in Unit 2 are found to
's' be different from those considered in the original Unit 1 qualification program.
The licensee has identified 17 and 34 equipment types in the Nuclear Steam Supply System (NSSS) and Balance of Plant (BOP), respectively, that are different.
1 The extent of the differences was judged to be of limited scope. Because of
{
this and the fact that the Unit I seismic qualification program was well implemented, the staff has determined that a plant site audit of Unit 2 equipment seismic qualification was not warranted.
For NSSS equipment the differences are due to changes in equipment design or location. For BOP equipment the differences are mainly due to design modifica-tions or procurement of similar equipment from different manufacturers. A complete list of equipment differences is provided in the submittai. For Unit 2 equipment found similar to Unit 1, qualification is based on and documented j
using Unit I qualification documentation and SQRT (Seismic Qualification Review
]
Team) forms. For equipment that is different, qualification was performed by i
developing rationale / calculations for location differences, providing evaluation j
for qualification using the new input spectra, reviewing qualification reports, and preparing SQRT forms for unique Unit 2 equipment. Plant modifications accomplished by Project Change Notices (PCNs) and affecting Unit 2 equipment qualification have also been evaluated.
For modifications similar to Unit 1 Limerick SSER 9 3-1
{
1
1 I
i - s modifications, Unit 1 evaluations are used as the basis. For modifications unique to Unit 2, calculations have been generated in support of Unit 2 i
3 qualification.
(
Finally, walkdowns were performed according to an approved procedure for Unit 2 f
equipment to ensure that installed equipment is consistent with qualification l
documentation. These walkdowns were performed on a sampling basis.
1 At the time of the March 7, 1989 submittals, all Unit 2 NSSS equipment and-about eighty-five percent of B0P equipment had been evaluated and determined to be acceptable for the required loading combinations. SQRT reports had been prepared for all items as is required. Preparation of the New Load Evaluations Final Summary Report was also in progress. These reports were completed and revised prior to fuel load and incorporated the results of the evaluation of discrepancies identified in the final walkdown. The licensee's evaluations indicated that none of the identified discrepancies was significant and did not affect the validity of the Unit 2 equipment seismic qualification.
Our review of the above information verified that the seismic and dynamic qualification of Unit 2 equipment was performed by an extension of the Unit 1 program. We conclude that the program as identified in Limerick Final Safety Analysis Report (FSAR) Sections 3.9 and 3.10 is applicable to both units, and the staff acceptance of the Unit 1 program as stated in the SER and its supple-ments is also valid for the Limerick 2 licensing application.
3.11 Environmental Qualification of Electrical Equipment Important to Safety and Safety-Related Mechanical Equipment 3.11.1 Introduction Equipment that is used to perform a necessary safety function must be demonstrated to be capable of maintaining functional operability under s
all service conditions postulated to occur during its installed life for N
the time it is required to operate. This requirement, which is embodied in General Design Criteria 1 and 4 of Appendix A and criteria III, XI, and XVII of Appendix B to 10 CFR 50, is applicable to equipment located inside as well as outside containment. More detailed requirements and guidance relating to the methods and procedures for demonstrating this capability for electrical equipment have been set forth in 10 CFR 50.49, " Environmental Qualification of Electrical Equipment Important to Safety for Nuclear Power Plants"; NUREG-0588,
" Interim Staff Position on Environmental Qualification of Safety-Related Electrical Equipment," which supplements IEEE Standard 323; and various NRC Guides and industry standards.
3.11.2
Background
NUREG-0588 was issued in December 1979 to promote a more orderly and systematic implementation of equipment qualification programs by industry and to provide guidance to the NRC staff for its use in ongoing licensing reviews. The positions contained in that report provide guidance on (1) how to establish environmental service conditions, (2) how to select methods that are considered appropriate for qualifying equipment in different areas of the plant, (3) other areas such as safety margin, aging and documentation for each item of safety-related Limerick SSER 9 3-2
t.
electrical equipment, and (4) to identify the degree to which their qualification programs complied with the staff positions discussed in NUREG-0588.
7
- IE Bulletin 79-01B. " Environmental Qualification of Class 1E Equipment," issued L
Janua ry.14, 1980, and its supplements dated February 29, September 30, and October 24, 1980, established environmental qualification requirements for operating reactors. This bulletin and its supplements were provided to OL applicants for consideration.
A final rule on environmental qualification of electrical equipment important to safety for nuclear power plants became effective on February 22, 1983. This rule, Section 50.49 of 10 CFR 50, specifies the requirements to be met for demonstrating the environmental _ qualification of electrical equipment important to safety located in a harsh environment.
In accordance with 10 CFR 50.49, electrical equipment for Limerick Generating Station, Unit 2 rey be qualified in accordance with the acceptance criteria specified in Category II of NUREG-0588.
In order to document the degree to which the environmental qualification program complies with the NRC's environmental qualification requirements and criteria, the applicant provided equipment qualification information by letters dated March 7, 1989, April 10, 1989 and Msy 9, 1989.
3.11.2.1 Purpose The purpose of the staff's review was to evaluate the adequacy of the Limerick Generating Station, Unit 2 environmental qualification program for safety-related mechanical equipment and electrical equipment important to safety as defined in 10 CFR 50.49.
3.11.2.2 Scope The scope of the staff's review was limited to an evaluation of the safety-related mechanical equipment and electrical equipment important to safety at N'
Limerick Unit 2 that is different from equipment at Unit 1.
This mechanical and electrical equipment must function in order to mitigate the consequences of a design basis accident, inside or outside containment, while subjected to the hostile environment associated with this type of accident.
Safety-related mechanical equipment and electrical equipment important to safety at Limerick, Unit 2 that are identical to equipment at Unit I were addressed in Supplement 2 (NUREG-0991).
3.11.3 Staff Evaluation By letters dated March 7, April 10, and May 9,1989, the applicant identified the following items of electrical equipment as specific Unit 2 equipment different from Unit 1.
Equipment Item Manufacturer Model Number 600V Power Cable Rockbestos XLPE Ins.
Flow Transmitter Rosemount 1153 Series B Differential Pressure Transmitter Rosemount 1153 Series B Limerick SSER 9 3-3
-Equipment Item-Manufacturer Model Number
! Level Transmitter Rosemount 1153 Series B j
Pressure Transmitter Rosemount 1153 Series B Motor Operator.
Limitorque SMB-00-10 Motor Operator
.Limitorque SMB-1-60 L
Pilot Solenoid Valve ASCO-206-832-3U-3RU i
Pilot Solenoid Valve
'ASCO NP8316A74E Electric Conduit Seal Patel Engineering Solenoid Valve Valcor V526-5000-Series Transformer Westinghouse 750 KVA q
The applicant also provided a summary description for extension of the Limerick Generation Station (LGS) Unit 1 mechanical equipment qualification
-(MEQ) program'(environment's) to Unit 2.
The summary was provided for staff review in order to (1) show that the MEQ licensing commitme es and program description, as identified in the LGS Final Safety Analysis Report (FSAR)
Section 3.11, is applicable to both units, (2) that the implementation for Unit 2 is an extension of the Unit 1 program, (3) identify minor differences in components or in items that affect qualification between Units 1 and 2 and-document the qualification acceptability of those differences, and (4) establish that the conclusion reached.in the SER for the acceptance of the LGS Unit 1 qualification program is also' valid for the application of the LGS program to Unit 2..
As a result of a review of.the information presented by the applicant,.
.the staff finds this approach to identification and qualification of both electrical.and mechanical equipment acceptable.
3.11.4 Conclusions The staff has reviewed the summary information provided by the applicant
.'e for the Limerick Unit 2 program for environmental qualification of. electrical equipment important to safety and safety-related mechanical equipment.
As noted above, this review is limited to Limerick Unit 2 safety-related
. mechanical equipment and electrical equipment important.to safety as defined in 10 CFR 50.49 that is different from equipment in Limerick, Unit 1.
The purpose of the review was to assess the qualification status of such equipment and to determine the adequacy of the qualification program.
Based on the results of our review and evaluation of the information provided by the applicant, the staff concludes that the applicant has demonstrated compliance with the requirements of 10 CFR 50.49, the relevant parts of General Design Criteria 1 and 4 of Appendix A, criteria III, XI and XVII of Appendix B to 10 CFR 50, and the criteria specified in NUREG-0588.
l Limerick SSER 9 3-4
?j e
s.
4 REACTOR 1
1 4.4 Thermal-Hydraulic Design 4.4.4 Thermal-Hydraulic Stability In Supplement'4 to the Staff's SER (NUREG-0991), the staff concluded that the Technical Specifications proposed for Limerick Unit I were consistent with the recommendations in General Electric Company Service Information Letter (CIL)-380 l
and acceptably resolve the thermal-hydraulic stability concern for Limerick Units 1 and 2, assuming long-term single-loop operation is not permitted. Should such operation be requested in the future, the staff will evaluate Limerick Units 1 and 2 Technical Specifications to determine if additional modifications are required.
On January 23, 1986, the staff issued Generic Letter No. 86-02, " Technical Resolution of Generic Issue B-19-Thermal Hydraulic Stability," On March 31, 1986, the staff issued Generic Letter No. 86-09, " Technical Resolution of Generic Issue No. B-59 (N-1) Loop Operation in BWRs and PWRs." On June 15, 1988, the staff issued NRC Bulletin No. 88-07; " Power Oscillations in Boiling Water Reactors" which requires adoption of certain operating procedures.
Supplement No. I to this Bulletin was issued December 30, 1988. The licensee responded to the Bulletin and Supplement I by letters dated September 5,1988 March 7, 1989 and March 31, 1989.
By application dated November 4,1988, the licensee requested approval of changes to the Unit I Technical Specifications (TS) to permit extended single loop operation. The licensee's letter of March 29, 1989 submitted the necessary analyses to support single loop operation of Unit 2.
By letter dated June 30,
,s'-
1989, the staff issued Amendment No. 30 to Facility Operating License No. NPF-39 for Unit 1.
The amendment and letter approved extended single loop operation for Limerick Units 1 and 2.
As part of our evaluation, we reviewed the operating restrictions proposed by the licensee in response to Bulletin 88-07 and Supplement 1.
As discussed in our letter of June 30, 1989 and the accompanying safety evaluation, we advised the licensee that the responses to Bulletin 88-07 and Supplement I satisfactorily resolved thermal hydraulic stability concerns for Limerick, Units I and 2.
In the letter of March 31, 1989, the licensee revised the response of March 7, 1989 to Bulletin 88-07, Supplement I to indicate that one exception will be taken to the implementation of the GE interim stability recommendations during the initial startup testing. As discussed in Chapter 14 of the revised Limerick FSAR, during Test Condition 4 Limerick Unit 2 will conduct recirculation pump trip testing which is specifically intended to identify any concerns with transients of the nature described in Bulletin 88-07 and any possible resulting instabilities. Because of the increased awareness of the possibility of instabilities and the increased monitoring by plant staff and augmented test instrumentation during the initial startup testing programs, the staff finds this one exception acceptable.
Limerick SSER 9 4-1
l 1
4 5
REACTOR COOLANT SYSTEMS 5.3 Reactor Vessel 5.3.1 Reactor Vessel Materials i
5.3.1.2-Fracture Toughness In the original SER, the NRC staff reviewed the fracture toughness ofLthe ferritic reactor vessel and reactor coolant pressure boundry (RCPB), and the materials surveillance program for the reactor vessel belt-line according to SRP.5.2.3.II.3.a and SRP 5.3.1.II.5, II.6, and II.7.
4 j
GDC 31 requires, in part, that the RCPB be designed with sufficient margin to ensure that when the boundary is stressed under operating, maintenance, testing, and anticipated transient conditions, it behaves in a nonbrittle 1
manner and the probability of rapidly propagating fracture is minimized.
GDC 32 requires, in part, that the RCPB be designed to permit an appropriate material surveillance program for the RCPB. Materials selection, toughness requirements, and extent of material testing were reviewed in accordance with the above criteria, subject to the rules and requirements of 10 CFR 50.55a, and Appendices G and H to 10 CFR Part 50.
At the time of the original review, the applicant had submitted only Unit 1 material information. Since that time, Unit 2 material information has been submitted both in.a letter on June 14, 1983, and in FSAR Revision 22.
However, on July 12, 1988, the NRC issued Generic Letter 88-11 "NRC Position on Radiation Embrittlement of Reactor Vessel Materials and Its Impact on Plant Operations." This Generic Letter identified tc licensees and applicants that N,
Revision 2 of Regulatory Guide 1.99, " Radiation Embrittlement of Reactor Vessel Materials," became effective in May 1988 and that it would be used by the NRC to review submittals regarding pressure temperature (P-T) that require an estimate of the embrittlement of reactor vessel beltline materials.
The licensee provided a response to Generic Letter 88-11 on November 23, 1988 that included the results of a Limerick Unit I fracture toughness analysis for the reactor vessel utilizing Revision 2 of Regulatory Guide 1.99.
These i
resu?ts concluded that i
1.
The Rev. 2 adjusted reference temperature (ART) values at 32 effective full power years (EFPY) for Unit I are below 200"F, which is the allowable limit in 10 CFR Part 50, Appendix G.
Therefore, implementation of Rev. 2 will not result in any additional analysis, testina or provisions for thermal annealing.
l l
l l
l 1
i i
Limerick SSER 9 5-1 A----
]
u s.
L 4
y
' 2..
TheARTvaluethatappliestothepressure-temperature (P-T)curvesin the Technical: Specifications is 56*F at 32 EFPY....The maximum Rev. 2 ART i
value: identified is 85.6*F.at 32 EFPY., Therefore..the Rev. 1 32 EFPY-P-T-L
- curves are less conservative than 32 EFPY curves that would be generated
!?
zwith Rev. 2.
However, the current P-T curves are applicable up to 10 1
EFPY if ART is calculated according to Rev. 2 methods. The licensee has committed to submit a. Technical. Specification amendment by December 29,.
1989 to specify.the revised time period for which the current curves are valid.
I 3.
The worst case low pressure coolant injection (LPCI) nozzle is also
' included in this beltline' region analysis due to its predicted neutron fluence value at 32 EFPY.: Since it has a Rev. 2 ART at 32 EFPY-that is less than the 40*F RT applicable to the limiting vessel discontinuity curves the LPCI nozzEis bounded by the limiting vessel discontinuity curves. Therefore, the discontinuity limits shown on the P-T curves of i
the Technical Specifications need not be adjusted as a result'of the
' implementation of Rev. -2 of Regulatory Guide 1.99.
The staff finds these conclusions and commitments acceptable until such time as. the review of the ' licensee's analysis is complete.
The licensee's November 23, 1988 response also indicated that a Unit 2 analyses would be submitted by April 28, 1989, and that any required Technical Specification revisions would be submitted by December 29, 1989. The applicant. submitted the Limerick Unit 2 analysis on March 31, 1989, which reached the following conclusions:
1.
The Rev. 2 ART values at 32 EFPY for Unit 2 are below 200*F, which is the allowable limit in 10 CFR 50, Appendix G.
Therefore, implementation of Rev. 2 will not result in any additional analysis, testing or provisions for thermal annealing.
2.
-The Rev. 1 ART value that applies to the beltline P-T curves (A', B',
and c') in the original draft Technical Specifications is 75'T at 32 EFPY. Therefore, these Rev. 1 32 EFPY P-T curves are less conservative than 32 EFPY curves'that would be generated with Rev. 2.
The applicant has generated new minimum reactor pressure vessel metal 4
temperature versus reactor vessel pressure curves for use in the Technical Specifications that are based on Rev. 2 of Regulatory Guide 1.99. Since these curves are more conservative than those based on Rev. 1, the staff finds their use acceptable until such time as the review of the licensee's analysis is complete. The' staff review is being conducted with the assistance of our Contractor, EG8G Idaho, and is nearly complete pending review of data on the nickel content on 11 beltline weld materials, weld surveillance material nickel content and other data.
I 5.3.2 Pressure-Temperature Limits This topic is discussed in Section 5.3.1.2 above.
Limerick SSER 9 5-2
l<
(.
i i
6 ENGINEERED SAFETY FEATURES 6.2 Containment Systems 6.2.5 Combustible Gas Control Inerting the containment for the LGS-2 plant is required by 10 CFR 50.44.
In 10 CFR 50.44, " Standards for Combustible Gas Control System in Light Water Cooled Power Reactors," Section 50.44(c)(3)(i) states in part that, " Effective May 4,1982 or 6 months after initial criticality, whichever is later, an inerted atmosphere shall be provided for each boiling light-water nuclear power reactor with a Mark I or Mark II type containment".
By letter dated December 5,(1988, the licensee requested an exemption from 10 CFR 50.44(c)(3) 1) to extend the permitted time of operation with a non-inerted containment to accommodate completion of the Power Ascension Test Program (PATP). The Limerick Unit 2 PATP is based on maintaining the containment in a non-inerted condition until the successful completion of the 100-hour warranty run, 5 condition that normally would be expected to occur within approximately 120 effective full power days of core burn-up.
The proposed exemption from the regulation is requested in order to complete the balance of the PATP in accordance with the licensee's test plan. No changes are being made in the maximum number of full power days of core burn-up normally expected before inerting is required. To assure this, the maximum expected value of 120 effective full power days is made part of the proposed action.
It is desirable to operate the reactor without inerting during the PATP, as an uninerted containment would permit unscheduled inspections or identification of possible problems during this period. The anticipated high frequency of N'
containment entries during the PATP period and the required deinerting and re-inerting time (about 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />) would tend to discourage early and frequent containment entries for identifying and correcting any potential safety problems.
Further, the NRC staff believes that to require inerting before the PATP tests have been completed could result in less assurance of safety, because of the added time and/or decreased sbility to directly examine and evaluate components and systems inside containment while the PATP tests are under way. Completing the PATP tests with an uninerted containment (exemption granted) would reduce the likelihood of development of an event requiring protective safety actions during the period of exemption. Because of a low fission product inventory during the PATP period, and the short duration anticipated for the exemption, there is an extremely low likelihood that the inerting system would be required.
Limerick SSER 9 6-1
i L
Based on the information provided by the licensee and experience at other BWRs,- the staff concludes that there will be no increase in risks of operation through completion of the PATP tests with the proposed limited exemption regarding initial inerting over the risks from postulated accidents with an inerted containment.
Therefore, since there is no perceived increase in risk by the mere fact of extending the time allowed for completion of the PATP tests under uninerted conditions, the NRC staff finds that operation would be as safe under the conditions proposed by the exemption as it would have been had the test been completed in the shorter calendar time of six months after initial testing.
Based on the considerations discussed above, we have concluded that the proposed temporary exemption from 10 CFR 50.44(c)(3)(1) is authorized by law, will not endanger life or property or the common defense and is otherwise in the public interest and should be granted.
l
.['
Limerick SSER 9 6-2
(-
q L..
j l-7 INSTRUMENTATION AND CONTROLS 7.4 Systems Reauired_for Safe Shutdown
]
7.4.2 Specific Findings 7.4.2.1 Capability for Safe Shutdown Following(Loss of Electrical Power to Instrumentation and Controls IEBulletin79-27) i Supplement 2 of the staff's Safety Evaluation Report related to the operation of Limerick Generating Station, Units 1 and 2 (NUREG-0991) evaluated PECO's various reports on the subject analysis. The reports included analysis of multiple control system failures due to High Energy Line Break (HELB) and common sensors instrument line or power source failures. The combined effects of an HELB and control system malfunctions were compared with the transient and accident analyses contained in the FSAR. The consequences-of the identified failures were bounded by the FSAR analyses, and the staff found the shared design configuration of certain power sources, sensors and instrument lines acceptable.
By letter dated February 17, 1989, PEco submitted a supplement to previous reports that identified areas where the current LGS Unit 2 design did not conform with the previously reviewed design. The impact of the current changes in Unit 2 design on the previous report's conclusions are determined by using
" comparative methodology" (Unit 2,1988 versus Unit 1,1983 plant design). By this method, failures in the Unit 2 equipment changes in the shared design configuration of certain power sources, sensors and instrument lines, and the consequential failure of multiple control systems is compared with the previously reviewed failures of Unit I shared' design configuration and the affected multiple control system.
In each case, the differences were found to be insignificant and bounded by the current LGS FSAR accident analysis.
The staff has evaluated the licensee's description of their assumptions, i
comparative methodology and determination that combined failure affects of HELB for the changed configurations with Unit 2 are bounded by the FSAR Chapter 15 analysis.
The submittal includes three sets of analyses to determine if a failure in a power source, sensor or instrument line or an HELB common to multiple Non-Safety Related Control System (NSCS) components will adversely affect the primary reactor parameters, i.e., vessel water level, pressure or reactivity.
4 Limerick SSER 9 7-1
7.4.2.1.1 Common Power Source Failure Analysis l
The current Unit 2 design of the station electrical distribution system has
)
minor differences in bus structures from that previously reviewed. The-differences have caused minor changes in bus loading and their combined failure effects. The changes are in 125 Vdc buses that are the preferred source of power to the instrument power supply inverters and several ac buses of different voltages. The licensee's analysis indicates that even with the changed config-uration of buses, the effects of a loss-of-bus event were not significantly different from those previously reviewed for Unit 1.
The critical buses chosen to fail for this analysis were those non-safety related buses that supply power to two or more major reactor non-safety related control system (NSCS) components.
i The effect of the loss of these strategic components on system operation and reactor primary parameters was insignificant and did not significantly deviate from Unit I analysis results.
7.4.2.1.2 Comon Sensor Failure Analysis Similar to Unit I design, the identified comon sensor lines and loads were mainly associated directly with the nuclear boiler applications. The sensor line support services both safety related system sensors and NSCS sensors.
Thus, a comon NSCS sensor line failure could cause changes in both the safety related and non-safety related actions. The Unit 2 nuclear boiler P& ids and various NSCS schematic diagrams were reviewed and the identified sensors were compared with Unit I comon sensor line loads. The differences were analyzed and found insignificant and bounded by the FSAR Chapter 15 analysis. For example, the reactor vessel taps, one each for lines 6 and 7, were reanalyzed after a difference was identified. A break in these lines will cause a reactor low water level (LWL) scram. A plug in these lines will inhibit a reactor LWL trip only by one set of channels. However, the redundant set of channels will not be affected by the plugged line and will trip the reactor when needed.
U 7.4.2.1.3 HELB and Affected NSCS Component HELB zone and the associated NSCS component are subject to the plant construc-tion and layout difference between Units 1 and 2.
The licensee performed a visual walkdown of the critical zones and component locations in Unit 2 similar to that performed at Unit 1.
The walkdown identified physical differences in some zone layouts and NSCS component contents between the Unit 1 zone layout and NSC component contents. The HELB failure analysis for such zones did not identify any new combined failure effects events. These events were similar to l
those in Unit 1 and were found bounded by FSAR Chapter 15 analysis.
Limerick SSER 9 7-2
l:
>c 7.4.2.1.4 Conc 1hsfons Based on the above_ evaluations, it is concluded that each of the three analyses adequately identified the differences between Unit 1 and Unit 2 L
electrical distribution systems. sensor lines, and HELB zone layout and NSCS E
component contents. The consequential common failure of control systems in
~
Unit 2 was adequately compared with those in Unit I to arrive at the conclusion that the failure events were bounded by the FSAR Chapter 15 analysis.
7.7 Control Systems 7.7.2 Specific Findings 7.7.2.1 High-Energy-Line-Break and Consequential Control Systems Failures (IE Information Notice 79-22) and Multiple Control Systems failures This topic is addressed in section 7.4.2.1.
i s
1 i
Limerick SSER 9 7-3 i
15 ACCIDENT. ANALYSIS 15.8 Anticipated Transients Without Scram 15.8.1 Introduction The original SER provided a review of the applicants action plans with regard to an anticipated transient without scram (ATWS) and found them acceptable, but indicated that the Commission would, through rule-making, determine any future modifications necessary to resolve the ATWS concerns.
On July 26, 1984, the Code of Federal Regulations (CFR) was amended to include Section 10 CFR 50.61, " Requirements for Reduction of Risk from Anticipated Transients Without Scram (ATWS) Events for Light-Water-Cooled Nuclear Power Plants" (known as the "ATWS Rule").
The ATWS Rule requires specific improve-ments in the design and operation of commercial nuclear power facilities to reduce the likelihood of failure to shutdown the reactor following anticipated transients, and to mitigate the consequences of an ATWS event.
For each boiling water reactor, three systems are required to mitigate the consequences of an ATWS event.
1.
It must have an alternate rod injection (ARI) system that is diverse (from the reactor trip system) from sensor output to the final actuation devices. The ARI system must have redundant scram air header exhaust valves. The ARI system must be desi reliable manner and be independent (gned to perform its function in a from the existing reactor trip system) from sensor output to the final actuation device.
U 2.
It must have a standby liquid control system (SLCS) with a minimum flow capacity and boron content equivalent in control capacity to 86 gallons per minute of 13 weight percent sodium pentaborate solution. The SLCS and its injection location must be designed to perform its function in a reliable manner.
3.
It must have equipment to trip the reactor coolant recirculating pumps automatically (recire pump trip or RPT) under conditions indicative of an ATWS. This elfuipment must be designed to perform its function in a reliable manner.
l This evaluation addresses the ARI system (Item 1), the SLCS (Item 2) and the l
ATWS/RPT system (Item 3).
Limerick SSER 9 15-1
4 15.8.2 Review Criteria The systems and equipment required by 10 CFR 50.62 do not have to meet all of the stringent requirements normally applied to safety-related equipment.
However, this equipment is part of the broader class of structures, systems, and amponents.important to safety defined in the introduction to 10 CFR 50, Appe6 dix A, General Design Criteria (GDC). GDC-1 requires that " structures, systems, and components important to safety shall be designed, fabricated, erected, and tested to quality standards commensurate with the importance of the safety functions to be performed." Generic Letter 85-06 " Quality Assurance Guidance for ATWS Equipment that is not Safety Related" details the quality. assurance that must be applied to this equipment.
'In general, the equipment to be installed in accordance with the ATWS Rule is required to be diverse from the existing RTS, and must be testable at power.
This equipment is intended to provide needed diversity (where only minimal diversity currently exists in the RTS) to reduce the potential for common mode failures that could result in an ATWS leading to unacceptable plant conditions.
The criteria used in evaluating the licensee's submittal include 10 CFR 50.62
" Rule Considerations Regarding Systems and Equipment Criteria" published in Federal Register Volume 49, No.124 dated June 26, 1984, and Generic Letter 85-06 " Quality Assurance Guidance for ATWS Equipment that is not Safety Related."
15.8.3 ARI & RPT System Description The Limerick Generating Station has installed a Redundant Reactivity Control System (RRCS) to mitigate the potential consequences of an anticipated transient without scram event. The RRCS consists of reactor pressure and reactor water les 11 sensors, logic,. power supplies, control room cabinets, and instrumentation tv initiate the protective actions to mitigate an ATWS event.
The protective actions include:
a.
Alternate Rod Injection (ARI),
b.
Recirculation Pump Trip (RpT),
c.
Feedwater Runback, and d.
Standby Liquid Control System (SLCS),
The RRCS is independent from the reactor trip system.
It is a two divisional l
safety related system.
Either division is capable of initiating protective I
actions when both input channels A and B within a division are tripped. The RRCS output will energize the devices to start the protective actions. The system can be manually initiated by depressing two push buttons (tripping both Channels A and B) in the same division.
1 i
The ARI logic will cause the immediate energiration of the alternate rod injection insert valves when either the reactor vessel high pressure trip setpoint or the low water level-2 trip setpoint is reached, or the manual Limerick SSER 9 15-2
4
.4 l
_ push buttons are armed and depressed. The AP.I valves and bleed paths are sized to allow insertion of all control rods to begin within 15 seconds. The j
status of the ARI system is indicated in the main control room.
The function of the RPT is to reduce the severity of thermal transients on fuel elements by tripping the recirculation pumps early in the transient events (such as turbine trip, or load rejections)., The rapid core flow reduction increases void content and.thereby introduces negative reactivity in the reactor to reduce the thermal power. There are two separate and independent.
.)
-(
systems-to trip the recirculation pumps. One is the reactor trip system end-of-cycle recirculation pump trip (EOC/RPT), which detects turbine control valve fast closure and main sto control system (ATWS/RPT)p valve closure. The other is the redundant reactivity which detects high reactor pressure or low reactor i
water levei. The Limerick design has two breakers in series for each reactor coolant recirculation pump. Each breaker has two independent trip coils; one receives a trip signal from the reactor trip system and the other receives a trip signal from the redundant reactivity control system. Both trip coils are Class IE qualified. The Class IE RTS and RRCS trip coils are separated from each other.. Each trip coil is capable of tripping the associated breaker independently of the other.
The RRCS detects high reactor pressure. After a 25 second time delay, it initiates the feedwater runback - provided the APRM (nuclear instrument average power range monitor) down-scale signal is not present. After a 100 second time delay, it isolates the reactor water cleanup system and automatically initiates the standby liquid control system.
The RRCS recirculation pump trip and feedwater runback are not initiated by manual initiation of the RRCS, However, these may be manually initiated at the respective system control panels.
The RRCS is continually checked by a solid state microprocessor based self-test system. This self-test system checks the RRCS sensors, logic, and N'
actuated devices. The RRCS sensors, logic and actuated devices and the APRM permitsive circuits are Class IE, independent of the RTS, and environmentally qualified. The ARI function can be reset by the ARI reset switches after a 30 second time delay to ensure that the ARI scram goes to completion. The other RRCS functions can be reset by the RRCS reset switches, provided the high reactor pressure or the low water level signal no longer exists.
15.8.4 Evaluation of ARI System 15.8.4.1 Safety Related Requirements (IEEE Standard-279)
The ATWS Rule does not require the ARI system to be safety grade, but the implementation must be such that the existing protection system continues to meet all applicable safety related criteria. The licensee stated that the ARI system (a subsystem of the RRCS) is classified as a Class IE system.
It is electrically diverse and independent from the reactor trip system, and it meets IEEE Standard 279-1971 in all applicable areas. The RRCS interfaces Limerick SSER 9 15-3
L' t
with control systems through the qualified isolation devices.
Any electrical failures in the control systems will not propagate into the PRCS to prevent ARI system from performing its protectiva functions. The staff finds this acceptable.
15.8.4.2 Redundancy The ATWS Rule requires that the ARI system must have redundant scram air header exhaust valves, but the ARI system itself does not need to be redundant.
Limerick's ARI system has redundant scram air header exhaust valves. Thr initiation and control circuits are redundant. All vent paths will allow insertion of all control rods to begin within 15 seconds and to be completed within 25 seconds.
The ARI performs a function redundant to the backup scram system. The staff finds this acceptable.
15.8.4.3 Diversity from Existing RTS The ATWS Rule requires the ARI system to be diverse from the existing reactor trip system. Limerick's ARI system uses energize-to-function valves instead-of deenergize-to-function valves.
It has DC powered valves and logic instead
'of AC powered valves and logic. Four reactor high pressure sensors and four low reactor reactor vessel water level sensors are dedicated for use to detect the ATWS events. The detection logic circuitries, power supplies and final actuated devices are independent from the reactor trip system. The built-in continuous self-testing feature will provide an additional assurance of reliability for the ARI system. The staff finds this acceptable.
15.8.4.4 Physical Separation from Existing RTS The ATWS Rule guidance states that the implementation of the ARI system must be such that separation criteria applied to the existing protection system are not violated.
The Limerick-ARI system sensors, transmitters, trip units and associated circuits are Class IE. The ARI system is separated and independent from the reactor trip system and has redundant divisions from sensor to the actuation of ARI valves. Either division can perform the protective action. The separation between two redundant divisions satisfies the guidance provided in Regulatory Guide 1.75.
The staff finds this acceptable.
15.8.4.5 Environmental Qualification The ATWS Rule guidance states that the qualification of the ARI system is for anticipated operational occurrences only, not for accidents.
The Limerick ARI system is a Class IE system.
It is qualified to the anticipated operational occurrence conditions. The staff finds this acceptable.
Limerick SSER 9 15-4
f a
L o..
15.8.4.6 Seismic Qualification No seismic qualification is required for ARI system hardware.
15.8.4.7 Quality Assurance NRC Generic Letter 85-06 dated April 16, 1985, provides quality assurance guidance for the-ARI system. The licensee is required to follow this guidance.
15.8.4.8 Safety Related (IE) Power Supply The ATWS Rule' guidance states that the ARI system must be capable of performing its safety functions with loss of offsite power, and that the power source should be. independent from the existing reactor trip system. The Limerick ARI systems are powered from Class IE 125 Yde power sources that are independent from existing reactor trip system power sources. Division I RRCS is powered by 125 Vdc from bus A. Division I.
Division'II RRCS is powered by 125 Yde from bus B, Division-II. These DC buses are backed up by station batteries. The staff finds that the ARI system is capable of performing its safety functions with loss of offsite power. The ARI power sources are independent from the existing RTS power source, and therefore this power supply arrangement is acceptable.
15.8.4.9 Testability at Power The_ ATWS Rule guidance states that the ARI system should be testable at power.
The Limerick ARI system is continually self-tested by a microcomputer based self-test system that tests the signal, trip setpoint and logic. An analog trip module (ATM) failure, out of calibration condition, or a. lack of system continuity condition will be annunciated. The ARI system uses a redundant 2-out-of-4 logic arrangement. Each reactor vessel level and pressure instrument can be tested during plant operation without initiating the ARI system, because two level or two pressure signals must be present in the same division to initiate the action. The staff finds-this acceptable.
15.8.4.10 Inadvertent Actuation The ATWS Rule guidance states that inadvertent ARI actuation that challenges other safety systems should be minimized.
The Limerick ARI system has redundant channels in each division. Both channels A and B must be tripped to initiate the protective actions. A manual initia-tion also requires arming the switch and depressing two push buttons to initiate the action. As a result, inadvertent actuation is minimized. The staff finds this acceptable.
15.8.4.11 Manual Initiation The Limerick ARI system has two sets of manual initiation switches (two switches in each division) in the control room. The operator must first rotate the Limerick SSER 9 15-5 i
j
' 4 push button collar to arm the switches, then depress both switches to initiate f
the protective actions. The staff finds this acceptable.
15.8.4.12 Information Readout The Limerick RRCS system provides status indications in the control room for potential ATWS, confirmed ATWS, ARI initiated, RRCS ready for reset and RRCS system related malfunctions. With continuous self-testing capability, the
~
operator always has current status of the RRCS. The staff finds this information presentation is adequate.
15.8.4.13 Completion of Protective Action Once Initiated The Limerick RRCS has a seal-in feature to ensure the completion of the protective action once initiated. After initial conditions return to normal, deliberate operator action is required to reset the safety system logic to normal. The staff finds this acceptable.
15.8.4.14 Maintenance Bypass There is no maintenance (manual) bypass of the RRCS. The staff finds this acceptable.
15.8.4.15 Conclusion Based on its review, the staff concludes that the ARI system design basis requirements identified above are in compliance with ATWS Rule 10 CFR 50.62 paragraph (C)(3) and guidance published in Federal Register Volume 49 No. 124 dated June 26, 1984, and is therefore acceptable.
l -
15.8.5 Evaluation of ATWS/RPT System 15.8.5.1 Safety Related Requirements The ATWS/RPT system is a subsystem of the RRCS and is classified as a Class IE system.
It is electrically diverse and independent from the reactor trip system, and meets IEEE Standards 279-1971 in all applicable areas. The staff finds this acceptable.
15.8.5.2 Redundancy The ATWS/RPT sy(end-of-cycle RPT).
stem has two trains. The ATWS/RPT function is redundant to the trip function The staff finds this acceptable.
15.8.5.3 Diversity from Existing RTS The ATWS/RPT system uses energize-to-function logic; instead of deenergize-to-function logic used in the RTS. The sensors, trip units, and power supplies of ATWS/RPT are diverse and independent from the RTS. The staff finds this acceptahle.
Limerick SSER 9 15-6
4 15.8.5.4 Physical Separation from Existing RTS The ATWS/RPT system sensors, transmitters, trip units and associated circuits are Class IE. They are separate and independent from the reactor trip system components. The staff finds this acceptable.
15.8.5.5 Environmental Qualification The ATWS/RPT system is a Class IE system.
It is qualified to anticipated operational occurrence conditions. The staff finds this acceptable.
15.8.5.6 Seismic Qualification No seismic qualification is required for the ATWS/RPT hardware.
15.8.5.7 Quality Assurance NRC Generic Letter 85-06 dated April 16, 1985, provides quality assurance guidance for the ATWS/RPT system.
The licensee is required to follow this guidance.
15.8.5.8 Safety Related (IE) Power Supply The ATWS/RPT system is powered from the Class IE 125 Yde power sources, which are independent from the existing reactor trip system. The DC' buses are backed up by station batteries; therefore, the ATWS/RPT system is capable of performing its safety functions with a loss of offsite power. The staff finds this acceptable.
15.8.5.9 Testability at Power The ATWS/RPT system uses a redundant 2-out-of-4 logic arrangement. Each level and pressure instrument can be tested during plant operation. The ATWS/RPT N'
system is continuously self-tested by a microcomputer based self-test system that tests the signal, trip setpoint and logic. An aralog trip module failure, an out-of-calibration condition, or a lack of system continuity condition will be annunciated. The staff finds this acceptable.
15.8.5.10 Inadvertent Actuation The ATWS/RPT system has redundant channels in each division. Both channels (A and B) must be tripped to initiate the protective actions. The ATWS/RPT actuation setpoint on reactor vessel high pressure is 1093 psig; the setpoint for reactor water low level is -38 inches. The RTS actuation setpoints on high reactor vessel pressure is 1037 psig; the setpoint for reactor water low level is 12.5 inches. Therefore, the ATWS/RPT actuation will not challenge j
the RTS. The staff finds this acceptable.
l L
Limerick SSER 9 15-7
h 15.8'5.11' Conclusion on ATWS RPT System o
i i
. Based on'its review. the staff concludes that the ATWS/RPT design basis'
.i requirements identified above r.re in' compliance with ATWS Rule 10 CFR 50.62
. paragraph:(C)(5) and guidanc.e. published in. Federal Register Volume 49 No.124 dated ~ June 26, 1984. The staff finds this acceptable.
'15.8.6 Evaluation of SLCS l
15.8.6.1.. Safety Related Requirements
. The standby liquid control system (SLCS) must have a minimum flow capacity and boron content equivalent'in control capacity to 86 gallons per minute of 13:
weight percent sodium pentaborate solution.
15.8.6.2 ' Evaluation The SLCS design information given' by the licensee has been reviewed by the; staff against the requirements.of the ATWS Rule (10 CFR' 50.62), and Generic
' Letter 85-03c" Clarification of Equivalent Control Capacity for Standby Liquid Control System," dated January 28,1985. The Limerick design.to meet the safety related requirement calls for two.of the three: installed SLCS pumps operating at a total combined flow rate of greater than or equal to 41.2 GPM, with the corresponding solution concentration, to meet the above equivalency.
r*quirement. Operation at this minimum requirement ryuires a solution concentration of not less'than 13.6%. The flowLeapacity and solution concen-tration provided by the licensee exceeds the ATWS Rule requirement of 86 GPM.
of 13 weigni. percent sodium pentaborate. This is acceptable.
The licensee's plan ic periodically test only one SLCS system pump at a time is also acceptable. This is based on the licensee's statement that tests performed on Limerick I during startup verified that the SLCS is capable of.
operating under the increased prescr_es associated with more than one pump
,j operation.
15.8.6.3 Conclusion on SLCS The license's design for the SLCS is acceptable, because it wili siiver an equivalent boron concentration of 13 weight percent sodium pentaborate at 86 GPM as required by 10 CFR 50.62.
l Limerick SSER 9 15-8 I
- 4 15.8.7
. Technical Specifications The equipment required by the ATWS Rule to reduce the risk associated with an ATWS event must be designed to perform its function in a reliable ma9ner. A method acceptable to the staff for demonstrating that the equipment satisfies the reliability requirements of the ATWS Rule is to provide equipment technical specifications, including operability and surveillance requirements.
The Limerick plant technical specifications have incorporated requirements for.the ATWS/RPT and 'ba.5LCS. The staff has not required technical specifications for the A:,' system.
l 15.8.8 Conclusions The staff has reviewed the design information provided by the licensee and concluded that the ARI design, the ATWS/RPT design and the SLCS design comply with the requirements of 10 CFR 50.62 and the guidance published in.the Federal Register on June.26, 1984 (49 FR 26036). The design of each of these systems is acceptable. Portions of this Safety Evaluation.were previously
' transmitted to the licensee on November 3,1987, and on June 8 (Amendment 22 to LGS-1). The staff has also reviewed the results of the surveillance tests performed by the licensee and has verified that the systems function as intended.
l
~
l l
L l
L Limerick SSER 9 15-9 l
l l
b
s
- 16. TECHNICAL SPECIFICATIONS Technical Specifications for Unit 2 were issued with the fuel load license and were re-issued unchanged with the low power license. These original Technical Specifications were developed to be identical to those of Unit 1, where possible, and amendments to the Unit 1 Technical Specifications were incorporated as discussed in SSER-8. Since the original issuance of the Unit 2 Technical Specifications, additional amendments were approved for Unit 1.
These amendments, listed in Table 16.1, and their safety evaluations are also applicable to Unit 2 and have been incorporated into the Unit 2 Tecnnical Specifications to be issued with the full power operating license. Some other minor typographical / editorial items have also been revised, but these items have no impact on the previous safety evaluations.
By letter dated May 31, 1989, Philadelphia Electric Company (PECo) (the licensee) provided a markup of the current Final Safety Analysis Report (FSAR) for the Limerick generating Station (LGS), Units 1 and 2.
The markup of FSAR pages was made to. incorporate the extended load line limit analysis (ELLLA), increased core flow (ICF) and partial feedwater heating (PFH) into the FSAR, so these modes of operation could be included in the draft Technical Specifications (TSs) for Unit 2.
To support these operational conditions and the draft revision to the FSAR, PECo'also provided two reports prepared by General Electric (GE). These reports are: NEDC-31577P " Extended Load Line Limit Analysis for Limerick Generating Station Unit 2. Cycle 1", dated March 1989, and NEDC-31578P " Increased Core Flow and Partial Feedwater Heating Analysis for Limerick Generating Station Unit 2, Cycle 1", dated March 1989 with Errata and Addenda No. 1, dated May 31, 1989. These markup changes were discussed with NRC staff on May 11, 1989 and'are consistent with the Final Draft version of the Unit 2 Technical Specifications transmitted by NRC letter dated May 19, 1989. The analyses supporting these modes of operation for Unit 2 are identical to the Unit I analyses that the NRC has previously accepted by safety evaluations dated February 17, 1987 (ICF and PFH), and August 14, 1987 (ELLLA).
N The PECo submittal proposes extensions to standard operating regions in the GESTAR II standard category of " Operating Flexibility or Margin Improvement Options". The selected options are ELLLA, ICF, final feedwater temperature reduction (FFWTR) and feedwater heater oct of service (FH005). These have become commonly selected and approved options for a number of reactors in recent years. These options are described and discussed in the GE topical reports for Limerick Unit 2, referenced above, which provide generic analyses of transients and accidents.
- The proposed ELLLA changes the Average Power Range Monitor (APRM) rod block and scram lines on the power-flow map, and permits operation along the new APRM red block line (0.58W + 50%) up to the intersection with the 100 percent power line, which occurs at a core flow of 87 percent. These dre standard changes for ELLLA. For ICF the approved flow increase is to 105 percent of rated core flow at 100 percent power. The increased flow is allowed throughout the cycle and after normal end-of-cycle (with or without FFWTR) with reactivity coast down. FFWTR involves feedwater temperature reduction up to 60%F (to 360"F at full power) and is proposed only for operation after a normal end-of-cycle. Limiting events have been analyzed for cycle extension to the exposure attainable using FH005, ICT and FFWTR at full power.
Limerick SSER 9 16-1
1 1
For the ELLLA extension, the topical reports discuss a full range of transient
'and accident events relevant to the region extension, and presents results of calculations or previously approved conclusions. The transient analyses demonstrate that the licensing basis results (e.g., 100 percent flow, 100 percent powe: for pressurization transients) bound the ELLLA region results (e.g., 87 percent flow, 100 percent' power). These conclusions apply to all relevant minimum critical power ratio (MCPR) events such as pressurization, rod withdrawal and flow runout events. Changes to MCpR TS are not required because of ELLLA adoption. Other relevant areas, such as overpressure protec-tion, LOCA and containment analysis have also been examined, and the analyses indicate that results are within allowable design limits. Thermal-hydraulic stability will be verified by appropriate surveillance. The analyses have been done with approved methodologies and the results are similar to previously approved ELLLA extensions..Thus, operation within the ELLLA region is acceptable for cycle 1 operation of Limerick Unit 2.
Nuclear transient data LOCA analyses and thermal hydraulic stability analyses consistent with the analyses previously performed for Unit I were developed to include the combination of ELLLA with PFH and ICF. Lower initial operating pressure and steam flow rate (due to lower feedwater temperature) provide more overpressure margin for the limiting MSIY closure flux scrom event.
Hence, it is concluded that pressure barrier integrity is maintained under PFH conditions. The licensee has analyzed the overpressurization limiting transient (MSIV closure) for increased core flow (ICF) without PFH. The analysis of this bounding transient predicted a peak sessel pressure of 1273 psig, which is below the ASME code limit of 1375 psig; the analysis results are therefore acceptable.
The fuel loading error accident, rod drop accident, and rod withdrawal error have been evaluated by the licensee for ICF and/or PFH operation. The rod withdrawal error transient is limited by a rod block system. The addition of a "high flow clamped" trip setpoint limit of 106 percent and allowable value of 109 percent of rated flow for the rod block monitor upscale alarm ensures N'
that the rod block trip value currently in the TS will not be exceeded. The reactor coolant system recirculation flow upscale trip setpoints and allowable values, and the values for the recirculation pump MG set scoop tube mechanical and electrical stops are increased. These changes are necessary to accommodate increase core flow operation and are acceptable. The licensee has stated that the fuel loading error and rod drop accident are not adversely affected by the proposed changes.
For the fuel loading error event, the licensee reported in their letter dated January 2,1987, a maximum increase in CPR of 0.04 from the value of 0.11 stated in the FSAR for this event at rated conditions. Thus the fuel loading error remains a non-limiting event. With regard to the rod drop accident, the LGS uses a banked position withdrawal sequence (BPWS) for control, rod movement. Based on prior staff review of BPWS as presented in Section S.2.5.1.3 of the General Electric Standard Application for Reactor Fuel (Supplement for US), May 1986 (NEDE-24011-P-A-8-US, as amended), the staff agrees that a fuel loading error event is not adversely affected by the proposed changes.
A loss of coolant accident (LOCA) with ICF and PFH was addressed in NEDC-31578P. The LOCA analyses with ICF alone bound operation with ICF and PFH. Since the peak clad temperature for ICF increases by less than 10*F for the limiting break compared to the rated core flow condition, the calculated Limerick SSER 9 16-2 l
peak clad temperature (PCT) of approximately 2100*F remains below the 10 CFR 50.46 cladding temperature limit. No changes to the current maximum average planar linear heat generation rates (MAPLHGR) are required.
In NEDC-31578P, GE stated that PCT changes throughout the remainder of the large break spectrum will be of a similar magnitude (less than 10*F).
Consideration was given to the break spectrum range of 60 to 100 percent DBA for the separate effects of ICF for several classes of BWR plants. The conclusion is that increased core flow results in a peak clad temperature increase of less than 10*F throughout the large break spectrum.
The separate effect of reduced feedwater temperature is to reduce the calculated peak clad temperature. A discussion was presented for both reduced t
feedwater temperature and increased core flow conditions, which bound the conditions described in the proposed amendment. Based on the staff's review of the information provided by the licensee, the staff agrees with the conclusion in NEDC-31578P that the effect of ICF will not alter the limiting break size. The calculated peak clad temperature remains below the 10 CFR 50.46 cladding temperature limits and is acceptable.
The impact of the proposed operating mode on containment LOCA response was considered by the licensee. A conservative analysis resulted in' a peak drywell deck downward differential pressure 2.6 psi higher than the value of 26.0 psid in the LGS FSAR. However, this is still below the design limit of 30.0 psid reported in the FSAR.
It was also stated that the peak suppression pool temperatures, chugging loads, condensation oscillations and pool swell bounding loads were all found to be bounded by the rated power analysis in FSAR Chapter 6.
We find this acceptable.
NEDC-31578P included a discussion of thermal-hydraulic stability (THS) for the LGS. The proposed LGS Unit 2 technical specifications implement a generic set of operating recommendations (Gereral Electric Service Information Letter No.
380, Revision 1, February 10,1984) to assure acceptable plant performance in the least stable portion of the power / flow map, and to provide operator instructions for the detect-and-suppress mode of operation. The THS compliance for all licensed GE BWR core fuel is demonstrated on a generic basis by NEDE-22277-P-1 and has been approved by the staff (NRC Safety Evaluation Report Approving Amendment B to NEDE-24011-P contained in Appendix US-C). PECo also committed in their letters of March 7, 1989 and March 31, 1987, to implement GE recommendations for thermal-hydraulic stability actions as outlined in NRC Bulletin No.88-07 supplement 1:
" Power Oscillations in Boiling Water Reactors (BWR)", dated December 30 1988. The staff concludes that acceptable THS provisions have been made.
fAlsoseeSection4.4.4)
We have reviewed the information provided by the Philadelphia Electric Company relative to the proposed operation of the Limerick Generating Station Unit 2 in the ELLLA region, combined with partial feedwater heating and increase core flow. Based on the results of the evaluation, the staff concludes that the preposed operations are acceptable. This information was previously transmitted to the licensee on June 14, 1989.
l Limerick SSER 9 16-3
c.
[,,..
.I i
Table 16.1 e
Additional Limerick Unit 1 Technical Specification Amendments
]
and Safety Evaluations Applicable to Limerick Unit 2
'l Original Amendment 4
Amend.
Submittal Issue No.
Date Date Subject 29 2/14/86 6/22/89 Clarification of.TS.
30 11/4/88 6/22/89 Single Loop Operation.
31-6/10/89 7/24/89 CRD Accumulator Testing.
S
d 1
l 17 QUALITY ASSURANCE 17.6 Readiness Verification Program In SSERs 7 & 8, the staff described the extensive Readiness Assessment and Readiness Verification Programs (RVP) being conducted by PEco to assess the i
design, construction and operational aspects of Limerick Unit 2.
A major feature of he RYP was an independent design and construction assessment (IDCA). The IDCA consisted of two major programs - an independent design j
assessment (IDA) and an independent construction assessment (ICA). PECo's performance of the IDCA and the NRC's inspection of the programs are complete.
The following provides the final status report on both programs.
17.6.1 Independent Construction Assessment (ICA)
In SSER-8,: the staff described the review of PEco's February 10, 1989 submittal and the follow up on-site inspection documented in Inspection Report 50-353/89-200. As' indicated, the team was favorably impressed with the
-licensee's efforts to determine the. scope of deficiencies identified by SWEC (Stone & Webster Engineering Corporation) and the NRC. However, uveral issues remained opened following the NRC inspection. The issues which required additional information from the licensee or additional review by the NRC are:
- 1) verification by the licensee that the wire size used for motor leads on the operator for valve HV-52-2F001C is adequate for its application, 2) a clarification by the licensee of its construction quality assurance program as it relates to the licensee's response to COR-056, 3) NRC review of additional information regarding resolution of grouted-in anchors that did not meet minimum embedment depths, and 4) an NRC-identified weakness associated with improperly performed quality control inspections.
On June 21, 1989, PECo responded to Inspection Report 50-353/89-200. PEco addressed the wire size issue in item 1 of Att;achment 1 and addressed both the clarification of its construction QA program and the QC inspections in item 2 of Attachment 2.
As indicated in the Inspection Report, the licensee provided the team with additional information regarding the grouted-in anchors immediately after the inspection.
As noted above, the licensee provided additional information about the wiring in the operator for valve HV-52-2F001C (core spray suction primary containment isolation) in the June 21, 1989 letter. The licensee stated in the letter I
that although the wire's continuous current rating is based on 30 degrees C and the maximum ambient temperature is 125 degrees F (51.7 degrees C), the wire insulation is rated for 125 degrees C.
In addition, the licensee indicated that the insulation is made of flame retardant cross-link polyethylene. The staff found that the motor lead wire used in the operator for valve HV-52-2F001C is adequate based on its insulation flame resistance and high temperature rating. The staff considers this open item to be adequately resolved.
In its June 21, 1989, letter, the licensee responded to items 2) and 4) above relating to quality assurance as one item, but the staff will discuss them separately here. The licensee described the Limerick quality program in detail and identified the various levels of reviews and inspections performed by the Limerick SSER 9 17-1
constructor (Bechtel)andthelicensee. The licensee also related this description of their construction quality program to Bechtel's references to the quality program, which the NRC found to be narrow, in Bechtel's response to COR 056. The staff found PECo's response to this open item in adequate and
{
considers this open item to be resolved.
Region I issued the final.open item as one of two violations in the Notice of Violations accompanying combined inspection report 50-352/89-10 and 50-353/89-16, which was issued June 30, 1989. The licensee responded to the Notice of Violations in its letter of July 28, 1989, and addressed the NRC's concern about a trend of inadequate QC aspections. The licensee acknowledged i
the violation from the NRC's ICA inspection with the clarification that'the examples cited by.the NRC were the result of improperly conducted inspections.
For corrective action, the licensee trained construction and quality control engineers to emphasize the importence of technical manual and drawing reviews before starting field installations and inspections.
In addition, the licensee reviewed the seven specific examples from the violation and determined that (1) no adverse trends existed, (2) the equipment would have functioned properly, (3) sufficient formal training was provided to personnel, and (4) Independent reviews of the plant's design and construction found the design to be adequate and construction to be of high quality. The staff found the licensee's response to the violation from the NRC ICA inspection to be adequate.
As discussed in Inspection Report 50-353/89-200, the inspectors found that some of the grouted-in anchors had embedment lengths less than specified on the design drawings. The staff has performed an engineering evaluation of the additional information provided by the applicant regarding safety significance of the deviation from the design specification. Two kinds of anchorage were used:
threaded rods without heads which were grouted in the concrete structure and used to support steel platforms and threaded bolts with nuts, ASME A-36 material, used as equipment anchorage. This evaluation addresses adequacy of both types of these anchors.
The staff reviewed Observation Report No. COR-034, Rev. O dated October 13, 1988 and the enclosed Non-Conformance Reports (NCRs) (NCR No. 14047 dated November 11, 1988 and eleven start-up NCRs dated between October 26, 1988 and November 14,1988). The staff also reviewed the supplemental information provided to the inspection team as noted in our letter of May 17, 1989. The letter indicated that the factor of safety for the grouted-in rods which were used for steel platforms, is so high that the reduction in capacity for pullout due to the less than the minimum required embedment is insignificant.
More specifically, a test program was undertaken in 1978 at the Limerick jobsite, to establish the shear and tension values for grouted-in anchors for various embedments.
(Limerick Generating Station
" Tensile Test Report on Grouted-in Anchors in Concrete Halls and Slabs," Bechtel Power Corporation, August 17, 1987 and Limerick Generating Station, Units 1 and 2, Job S031,
" Shear Load Testing of Grouted-in Anchors, Specification 8031 C-51," dated September 21,1978). The results indicate that for 1 inch diameter anchors j
with 6 inch embedment the shear capacity was 20.8 kips. Similar results were obtained for other size anchors. The specimens were threaded rods and the compressive strength of concrete was 5.0 ksi. Since the concrete strength at Limerick was 4.0 ksi, the corresponding shear value was reduced accordingly, in Limerick SSER 9 17-2
~
1
-z proportion to the square root of the respective concrete strengths. This was
)
compared with the allowable value of approximately 2 kips, which would have an ample factor of safety against the test results. Through discussions with 1
Bechtel and PECo personnel, the staff established that the tests reflected the l-installation conditions at the site.
1 In order to assess the safety significance of the reduced embedments of the grouted-in anchors which were used for equipment supports a statistical j
analysis was performed.
i In has been determined that 919 grouted-in anchors have been installed in Unit 2 to date on Seismic Category I large pipe supports. A sample, consisting of 136 (15 percent) of the installed grouted-in anchors was examined. Out of the l
136 specimens, 31 rods were found to not meet the minimum embedment l
requirements as specified on the pertinent drawings. At the request of the a
staff the applicant submitted documentation which reports on the statistical l
analysis.
(Calculation for Capacity of Grout-in Rods with Reduced Embedment, 1
Calculation #114.11.13, Rev. 1.) The sample of 136 anchors was tabulated and f
separated into populations by diameter. The quality of each population was specified by computing a lower bound, called tolerance limits. These tolerance limits were calculated so that 95 percent of the population should fall above the limit at a 95 percent confidence level, i.e., the probability that 95 percent of a population falls above the tolerance limit is 0.95.
By comparing the tolerance limits and the actual embedment lengths it can be established that out of the sample of 136 anchors examined two were outside of tolerance limits (one - 5/8 in. diameter and one 3/4 in. diameter). We consider this to be acceptable.
This statistical evaluation was used as the basis for determination of the design allowable values of the threaded rods.
The design loads for the anchors which did not meet the minimum embedment requirements were compared with the corresponding design allowable values using the interaction formula. The results of this comparison was reviewed by the staff and found acceptable.
A question was raised in connection with the applicant's analysis of seismic qualification of steel platforms using the grouted-in rods as anchorages. The loads on grouted-in rods were obtained from the equivalent static analysis using peak acceleration but without the factor of 1.5 as required by the Standard Review Plan Section 3.7.
These loads were compared with those obtained by the dynamic analysis using the response spectrum analysis technique and the BSAP (Bechtel Structural Analysis Program) computer code. The reactions on grouted-in rods obtained from the equivalent static analysis described above were compared with the corresponding loads obtained from the c'ynamic analysis. Review of the results indicates that the equivalent static analysis without the factor of 1.5 provides a conservative estimate as compared with ti:e actual loads obtained from the dynamic analysis.
It is therefore concluded that the use of equivalent static analysis without the 1.5 factor is acceptable for determination of the seismic loads on the platforms.
On the basis of the review described above, the staff concluded that the deficiency in embedment of the grouted-in rods as observed during the special i
inspection at Limerick Unit 2, conducted in March 1989, is not significant 1
Limerick SSER 9 17-3 i
t-s.
enough to cause a risk to public safety and that-the construction can therefore be considered as' acceptable. - The above conclusion is based on the following:
(a)' The> applicant conducted tests _ on similar type of grouted-in rods with and without nuts and the test results indicate that there is an ample factor of. safety against the allowable loads, indicating considerable conservatism
' inherent in the design of the rods.
(b) The' steel platforms supported by.the grouted-in rods were investigated by computerized dynamic analysis as well as by the equivalent static analysis and the results compared. The loads in the analysis from both were computed using the straight line interaction equations'and the results areislightly more conservative than those obtained by the-computer analysis.
(c) A statistical analysis of the sampled embedded rod length data was made which indicated that out of the 136 grouted-in rods only two are outside of 95 percent tolerance limits. We consider that this is insignificant.
Based on the above the staff concludes that for the rods / anchors that did not meet the minimum embedment depths specified on the drawings that:
the deviations was not sufficient to present a concern with respect to integrity of the construction of the plant or to cause a' concern regarding their intended functions and the as-found condition of the grouted-in rods / anchors is acceptable.
The staff has reviewed the above information and the additional information provi.ded in Attachment 2 to the licensee's June 21, 1989 submittal. The staff-finds these responses and the licensee's May 16, 1989 confirmation that the action items resulting from the construction portion of the 10CA cre complete as sufficient-documentation to close the independent construction assessment
,s review of Limerick Unit 2.
The team concludes that the Limerick Unit 2 IDCA and the staff's reviews have confirmed that the Limerick Unit 2 construction program has been satisfactorily implemented.
17.6.2 IndependentDesignAssessment(IDA)
In SSER-8, the staff described the review of PECo's April 12, 1989 submittal and the follow-up inspection conducted the week of April 24, 1989.
(Inspection
-Report 50-353/89-201, notyetissue). As indicated, the team concluded that the IDA provided the needed additional design assurance that Limerick Unit 2 has met its licensing commitments. However, this was contingent upon PECo's providing acceptable responses to six (6) items. As stated in SSER-8, two of the six items were satisfactorily resolved, except for the supplement hszards report. The licensee submitted the Hazards Program Evaluation Supplement on 4
May 25, 1989, and included additional information which provided acceptable confirmations of the actions requested in item (6), SSER-8. The Hazards i
Program and SWEC's evaluation indicated that the program and its implementation
]
are acceptable. The staff concludes that the Limerick Unit 2 IDCA and the i
staff's reviews have confirmed'that the Limerick Unit 2 design program has been satisfactorily implemented.
1 Limerick SSER 9 17-4
A 17.6.3'
' Conclusions PEco has initiated and completed on independent assessment of the design and
-j construction processes for Limerick Generating Station Unit.2. The staff has
'l reviewed this assessment and found it to be appropriately conducted and its i
results valid. The results conclude:
1 The design of safety related systems and structures for Limerick 2 complies with licensing commitments.and is technically adequate.
The construction of' safety related systems and structures for Limerick 2 is satisfactory and is generally in accordance with drawings -and
-specifications.
i The. design and construction process employed-for Limerick 2 is an I
acceptable process.
4 i
l Limerick SSER 9 17-5
.s APPENDIX A CHRONOLOGY LIMERICK GENERATING STATION, UNIT 2 May 31, 1989 Letter from applicant certifying that facility designed, constructed and tested in compliance with 10 CFR 20, 50, 51 and 100 with exception of specific exemptions requested per Commission regulations, Section 50.12 and Commission May 5,1989 order.
May 31, 1989 Letter from applicant forwarding marked-up FSAR pages, incorporating extended load line region, increased core flow, partial feedwat2r heating and proprietary GE Reports NEDC-31577P and NEDC-31578P.
Reports withheld (reference 10 CFR 2.790).
May 31, 1989 Letter to licensee forwarding Amendment 21 to License NPF-39 and safety evaluation. Amendment revises Technical Specifications to permit use of increased pore size of filters used during testing of diesel generator fuel oil.
June 2, 1989 Letter from applicant responding to NRC May 19, 1989 request for utility to review and certify final draft version of Unit 2 Technical Specifications. One major change that is required to be made to final draft is incorporation of changes to standby liquid control system Tectnical Specifications.
s June 2, 1989 Letter from applicant submitting follow-up activity in which four fuel assemblies containing fuel rods with alternate clad surface processes will be inserted into facility initial core. No new materials introduced to reactor environment.
June 2, 1989 Letter from applicant confirming that rod sequence control system deleted and rod worth minimizer setpoins lowered, per NRC February 7,1989 approval cf November 9, 1988 request for subject changes.
June 2, 1989 Letter from applicant forwarding Revision 1 to "PECO Response to NRC Bulletir,88-005 for Limerick Generating Station Unit 2," incorporating comments and providing additional information per April 27, 1989 and May 18, 1989 requests.
Limerick SSER 9 Appendix A
June 5, 1989 Letter from licensee forwarding applicant motion for clarification of Comission delegation of authority and for issuance of operating license or alternatively, for exemption from any requirement that license for Unit 2 cannot be issued.
June 5, 1989 Letter from applicant requesting that NRC be prepared-to issue license authorizing unit fuel loading and operation up to 5% of rated power as early as June 16, 1989. Testing activities that may not be. completed at-initial fuel loading enclosed.
June 7, 1989 Letter from licensee forwarding signed affidavit of C. A. McNeill, per applicant June 5, 1989 letter to be-substituted for copy attached to applicant motion, for clarification.of Commission delegation of authority and for issuance of operating license or for exemption.
June 8, 1989 Letter to licensee forwarding Amendment 22 to License NPF-39 and safety evaluation. Amendment revises Technical Specifications regarding standby liquid control system to ensure compliance with Paragraph (c)(4) of ATWS rule (10 CFR 50.62) to simplify and improve specifications for system.
June 9,'1989 Letter from licensee requesting temporary waiver'of compliance for facility to allow time for processing of emergency Technical Specification change request.
Depressurization of accumulators through accumulator
.N check valves not significant in current operating condition.
June 10, 1989 Application for amendment to License NFF-39, consisting of Technical Specification Change Request 89-05, revising Surveillance Requirement 4.1.3.5.b.2 regarding control rod accumulator check valve measuring and recording times.
June 14, 1989 Letter to licensee forwarding Amendment 23 to License NPF-39 and safety evaluation. Amendmut changes Technical Specifications to reflect completion and tie-in of standby gas treatment system and refueling area HVAC systein.
June 15, 1989 Letter to licensee forwarding Amendment 24 to License NPF-39 and safety evaluation. Amendment revises Technical Specifications to reflect incorporation of Unit 2 power supplies needed to support common equipment used in operation of Unit 1.
Limerick SSER 9 Appendix A L
June 14, 1989 Letter from licensee advising that plant operations review connittee chairman will be responsible for ensuring that individual who satG 'es Regulatory -
Guide 1.8, Revision 1-R qualificatw.s for manager will participate in connittee activities, per May 17,
'1989 discussion.
June 14, 1929.
Letter to licensee forwarding safety evaluation regarding increased core flow analysis, partial feedwater heating analysis and extended load line limit analysis.
June 15, 1989 Letter from applicant confirming that utility will be ready for NRC to issue operating license.to permit fuel load to begin on June 21, 1989, per 10 CFR 50.50, which permits ir.itial loading of fuel to begin on June 21, 1989 and provides clarification of information contained in June 5,1989 letter.
June 16, 1989 Letter to licensee forwarding Amendment 25 to License NPF-39 and safety evaluation. Amendment revises Technical Specifications to increase minimum level of water that must be maintained in spray pond to support operation of Unit 2.
June 19, 1989 Letter from licensee confirming (withdrawal of request for exemption from 10 CFR 50.44 c)(3)(ii)(8).
l June 19, 1989 Letter from licensee providing status of
,N modifications regarding four outstanding control room l
human engineering discrepancies (HEDs) identified l
during Unit 1 CRDR program. HED TA-03 resolved and l
HED Al-02 reevaluated and priority lowered from priority 2 to 4.
June 19, 1989 Letter to licensee forwarding Amendment 26 to License NPF-39 and safety evaluation. Amendment revises effluent dose limits to per site basis.
June 20, 1989 Letter to licensee forwarding Amendment 28 to License NPF-39 and safety evaluation. Amendment revises Technical Specifications to delete requirements that APRMs be operable when plant is in cold shutdown condition.
Limerick SSER 9 Appendix A l
L.
l c
June 20, 1989 Letter to licensee forwarding Amendment 27 to License NPF-39 and safety. evaluation. Amendment revises Technical Specifications regarding RHR and emergency service water system to reflect unit operation.
June 20, 1989 Letter to applicant forwarding Safety Evaluation Report (SER) accepting utility March 31, 1989 and June 2, 1989 responses to NRC Bulletin 88-005,
" Nonconforming Materials Supplied by Piping Supplies, Inc. at Folsom, NJ and West Jersey Mfg Co. at Williamstown, NJJ June 20,1989 Letter to applicant approving utility June 21, 1988 power ascension test program for facility, per Regulatory Guide 1.68, for execution.
June 21 1989 Letter to licensee forwarding SER accepting utility-Septem' er 6,1983, November 10, 1983 and May 8, 1984 J
responses to Generic Letter 83-28, Item 4.5.3 regarding on-line functional testing of reactor trip system, per BWR Owners Group Report NECD-30844.
June 21, 1989 Letter from applicant forwarding response to items requiring additional information and revised or supplemental information to Independent Construction i
Assessment Inspection Report 50-353/89-200.
June 22, 1989 Letter to applicant forwarding License NPF-83 authorizing fuel loading and precriticality testing of utility and FR notice. Amendment 4 to Indemnity Agreement B-101 also enclosed.
June 22, 1989 Sumary of June 15, 1989 meeting with Office of Executive Director, Nuclear Reactor Regulation, Office of General Counsel and Office of Nuclear Regulatory Research regarding progress of work on severe accident mitigation design alternative issue for plant.
June 22, 1989 Letter to licensee forwarding Amendment 29 to License NPF-39 and safety evaluation. Amendment revises Technical Specifications to achieve consistency, remove outdated material and make minor text changes and correct errors.
Limerick SSER 9 Appendix A
1 a=
.l I
. June 23 1989 Letter from licensee 'adyising that utility did not purchase any safety-related components and/or parts for Planned Maintenance System, Inc. from July.1, 1985 to present.
' June 23, 1989 Letter from licensee responding to request for additional information regarding consideration of severe accident mitigation design alternatives.
Tables listing current estimated core. damage frequency per reactor year and dominant population dose sequences enclosed.
June 28, 1989 Generic Letter 89-10 to all licensees of operating nuclear power plants and holders of cps for nuclear power plants regarding safety-related motor-operated valve testing and surveillance.
' June 28, 1989 Letter to licensee forwarding corrected Page 3/4 7-5 and overleaf Page 3/4 7-6 for Technical Specifications to increase minimum level of water that must be maintained in spray pond to support operation of units per Amendment 25 to license NPF-39.
June 28, 1989 Letter to licensee forwarding corrected Pages 3/4 3-21 and 3/4 3-22 for Amendment 29 to License NPF-39.
June 29, 1989 Letter from licensee; forwarding Amendment 95 to OL application for Licenses NPF-39 and NPF-83, consisting of Revision 58 to FSAR (filed in Category K) and Revision 12 to fire protection evaluation
^
i>
report and Revision 19 to emergency plan (filed in Category F).
June 29, 1989 Letter from licensee forwarding Revision 58 to FSAR (filed in PDR Category K).
June 29, 1989 Letter from licensee forwarding Revision 12 to fire protection evaluation report and Revision 19 to emergency plan (filed in PDR Category F).
Limerick SSER 9 Appendix A
'<g.
't
a-b l..
June 30, 1989~
Generic Letter 89-11 to all. holders'of OLs or' cps for -
BWRs regarding resolution of. Generic Issue 101, '.'BWR -
Water Level Redundancy."
July 3,1989.
Letter to applicant forwarding Technical Specifications, NUREG-1360 and SSER 8 regarding application for OL for Unit 2.
Without enclosure.
Limerick SSER 9 Appendix A
F
- y-
-O APPENDIX H Principal Staff Contributors Supplement 9 to the SER is a product of the NRC staff. The NRC staff members listed below were principal contributors to this report.
Name Unit Iqbal Ahmed Instrumentation & Controls 1:
Walter R.. Butler Project Directorate I-2 Richard J. Clark Project Directorate I-E Kulin D..Desai Reactor Systems Michele Evans Resident Inspector Roy L. Fuhrmeister Resident Inspector Mark Hartzman Mechanical Engineering Thomas J. Kenny Senior Resident Inspector Hulbert C. Li Instrumentation & Controls Margaret B. O'Brien Project Directorate I-2 Ronald W. Parkhill
.Special Inspections Howard J. Richings Reactor Systems Larry L. Scholl Resident Inspector Carl S. Schulten Technical Specifications Steven R. Stein Special Inspections George Thomas Reactor Systems Ed Trottier Project Directorate I-2 John C. Tsao Materials Engineering
.lr Limerick SSER 9
-I-Appendix H
r.,.
~
e o
APPENDIX U ERRATA to the Safety Evaluation Report for the Limerick Generating Station Location Current Wording Revision Pg vii, line 15 "other' "Other" line 25/26 Add "H PRINCIPAL STAFF CONTRIBUTORS" Pg 1-1, line 6
" licensees"
" licenses" line 42 "Public" "Public" Pg 1-2, line 24 Add Stairway"
" Add Stairway" line 28
" Reactor"
" Rod" Pg 3-1, line 35 "P, is" "P
is" Pg 3-2,'line 32
"(fSInc.)"
"kSI)"
line 33
"(WJMC)"
" WJM)"
line 34
"(CLMM)"
" CLM)"
line 39 "PS Inc., WJMC,
" PSI, WJM and CLM" and CLMM" Pg 4-2, line 8
"(Ref. 3)"
delete Pg 6-1, line 4 "SGRS" "SGTS" line 27 "GDC-56; and" "GDC-56 and" Pg 7-1, line 4 "7.2.27" "7.2.2.7" line 7 "Amendnment 45"
" Amendment 45" Pg 7-2, line 27 "NEDE-24011-P A" "NEDE-24011-P-A"
~
Pg 7-3, line 6
" bypassing"
" Bypassing" line 15
" staff"
" staff's" Pg 10-1, line 13/14 Add blank line Pg 15-1, line 19 "2.1.1. and" "2.1.1 and" Pg 16-2, line 32/33 "Diffe "/"rential"
" Differ "/"ential" Pg 17-1, line 22
" Limerick Unit 1"
" Limerick Unit 2" line 29
" function An"
" function. An" Pg 17-3, line-10 "commtments"
" commitments" line 15 "supplemtnal"
" Supplemental" Pg 18-1, line 5
" Nuclear" delete Pg 18-2, line 29 "an displays" "and displays" Pg 18-3, line 26
" Nuclear" delete Pg 19-1, line 35
" operating"
" operation" Pg 22-1, line 8 "50.33(k)(a)"
'"50.33( k)(2)"
Pg 22-2, line 5
"(20 in"
"(2)in" line 8 "50.12(a)(v),"
"50.12(a)(2)(v),"
Appendix H missing Included with SSER-9 Appendix U Pg 1, line 18 "100 percent" "100-percent" line 25 "Page 20-3,"
"Page 10-3,"
Limerick SSER 9 Appendix U