ML20235X375
| ML20235X375 | |
| Person / Time | |
|---|---|
| Site: | Pilgrim |
| Issue date: | 07/21/1987 |
| From: | Varga S Office of Nuclear Reactor Regulation |
| To: | Kane W, Shao L NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I), Office of Nuclear Reactor Regulation |
| Shared Package | |
| ML20234D382 | List: |
| References | |
| FOIA-87-643 TAC-65356, NUDOCS 8707240158 | |
| Download: ML20235X375 (14) | |
Text
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MEMORAE00M FCU ladenceShao, Director y Division of Engineering and Systems Technology William F. Kane, Director i
DivislonofReactorProjects,'RI-FROM:
Steven 3. V$rga, Director
_m D 1 Division of Peactor Projects I/II
SUBJECT:
INITIAL ASSESSMENT OF FILGRIM St.FETY ENHANCEMENT PROGPAM On July 8,191}7, Boston Edison Company (BEco) submitted a detailed description of the Pilgrim Safety Enhancemnt Program (SEP.) to the NRC.
(Copies of this submittal have been provided to you separately; however, contents of this submittel are summarized in Enciow re li. The submittal describes hardware changes that BECo has voluntarily elected te inplement at Pilgrim. BECo states in their submittal that une of the dysical plant changes increases the probability or consequences of a design basis accident and that all of the chahges wjll result in a reduction 10 the frequency of core melt scenarios or j
(c im; covement in tne performance 1$the containment response. REco h 4 i
'ad7ised the PM that all of the chups could be implemented under the
- hbvisions of 10 CFR 50.59.
7 s
BECb hahadvised the staff that thev.dytend to implement thed changey prior i
to' restart of the Pilgrim facility lestimated by BFCo to be in late,
3 C*,ptember). In a July 16lG87 conversation with Ralph Bird (AEco Senior d
Vice-President - Nuclear), Dr. Murley committed to a prompt staff assessment
~
of these changes to determine their safety impact anhto evaluate the
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ifccr.see's approach to their implementation. As a part of our initial
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avarssment of these changes, a visit to the BECo engineering o' ffices in Brcintree, MA, is planned for July 22, 1987. Note thaOwe expect to expend additional review effort subsequent to our return and that Region I plans j
additional inspections of SEP modification activities at the site.
I have directed the Pilgrim PM (Dick Wessman1 to lead a multi-disciplined team including both NRR and Region I personnel to make this Visit.' Suggested representatives are identified in Enclosure 2.
l To structure the team's effort and to allow me to report the results of this initial assessment promptly to Dr. Murley and the utility # the guidelines and' l
summary report format cA Enclosures 3 and 4 shou 11 be followed. Also includsd for information'(s guidar,cegegarding 10 CyR 50 9 reviews extracted from the
< s IE ManuaP (Enclo;ure 5).
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,, July 21, 1987 I appreciate your "short-fused" support on this effort.
Please contact B. Boger (X27415) or D. Wessman (X24937) if you have questions.
TAC No. 65356 is assigned to this effort.
Steven A. Varga, Director Division of Reactor Projects I/II L
)
Enclosures:
As stated cc:
T. Murley J. Sniezek F. Miraglia R. Starostecki A Thadani F. Rosa J. Craig W. Hodges J. Joyce J. Wiggins, RI C. Tinkler N. Su V. Thomas
- 0. Chopra L. Briggs, RI DISTRIBUTION Docket Files PD I-3 R/F SAVarga BABoger VNerses RHWessman MRushbrook DRPR:PDI-3 DRPR: PDI-3 DRPR:AD RI D
RHWe sman:ah VNerges BABoger a
a 07
/87 07/&U/87 074AU/87 07
/87 l
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I appreciate your "short-fused" support on this effort. Please contact B. Boger (X27415) or D. Wessman IX 4937) if you have/4uestions. TAC No. 65356 is assigned to this effort.
Ste A. Varga, Director Di is n of Reactor Projects I/II
Enclosures:
As stated cc:
T. Murley l
F. Miraglia R. Starostecki A. Thadani F. Rosa J. Craig W. Hodges J. Joyce J. Wiggins, RI C. Tinkler
'N.
Su V. Thomas
- 0. Chopra L. Briggs, RI DISTRIBUTION Docket Files PD I-3 R/F SAVarga BABoger YNerses P.HWessman FRushbrook
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RPR:PDI-3 POI-3
. D RI DRPR:DIR P.HWessman:a h es ABoger SAVarga 07/ c/87 07/js/87 07/0/87 07/ /87 7
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' l.0 INTRODUCTION l
1.1 Purpose of Report 1.2 Scope of Report 1.3 Safety Enhancement Program Goals 1.4 Safety Enhancement Program Plant and Operational' Changes 2.0 OVERVIEH OF SAFETY ENHANCEMENT PROGRAM
2.1 Background
l 2.2 Safety Enhancements
)
l
3.0 DESCRIPTION
OF SPECIFIC PLANT SAFETY ENHANCEMENTS 1
3.1 General Considerations 3.2 Installation of a Direct Torus Vent System (DTVS) 3.3 Containment Spray Header Nozzles 3.4 Additional Sources of Hater for RPV Injection and Containment Spray 3.5 Diesel Fire Pump for RPV Injection and Containment Spray 3.6. Diesel Pump Fire Pump Fuel Oil Transfer System f
3.7 Backup Nitrogen Supply System 3.8 Blackout Diesel Generator Including Protected Installation Facilities 3.9 Automatic Depressurization System Logic Modifications 3.10 Addition of Enriched Boron to Standby Liquid Control System 3.11 ATHS Feedwater Pump Trip
]
3.12 Modifications to Reactor Core Isolation Cooling System Turbine Exhaust Trip Setpoint 3.13 Additional ATHS Recirculation Pump Trip
4.0 DESCRIPTION
OF OPERATIONAL PLANT SAFETY ENHANCEMENTS
5.0 CONCLUSION
S
ENCLOSURE 2 INITIAL ASSESSMENT TEAM MEMBERS FUNCTIONAL AREA ORGANIZATION MEMRER Panagement/ Coordination PD I-3/NRR R. Wessman Plant Systems SPLB/NRR C. Tinkler Peactor Systems SRXB/NRR N. Su Instrumentation SICB/NRR V. Thomas Electrical SELB/NRR
- 0. Chopra i
Implementation /50.59 Application DRS/RI L. Briggs G
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2 a
lNITIAL ASSESMENT GttIDELIt:ES j
The following are suggested guidelines for use in conduct of the initial assessment.
Any conclusions reached about the technical adeauscy or method of implementation of the FECo SEP modifications are considered tentative.
This assessmnnt 3s brint and cannot reflect en indepth technical review, due to the constraints of time.
Each SEP enhancement thould be essrssed with consideration of the following:
1.
What is the safety impact of the change when considerrd alone or along with the other changes?
Does an "unreviewed safety question" exist?
(Criteria for determining whethor an unrevicted safety question exists are defined in Peregraph (a)(2) of 10 CFR 50.59.
Copy attached).
2.
Is e change to the Technical Specifications ren,uired?
(If the ansuce is "Yes" the modification and the proposed Technical Specification change must be reviewed by the staff before implementation).
3.
For those items in which no Technical Specificat. on change is i
involved, should the licensee be allowed to implement the change before staff review is complete?
4.
Assess the adequacy of the licensee's evaluation And conclusions regarding each SEP item.
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a a StiMMARY _ PEPnRT
~ '%T For each SEP item you assess, provide a brinf silmc ar y repor t using the following format.
This is to facilitate management decisionmal:2ng and assure consistency in app.oach.
Hopefu))y, each summary won't require more than 1-2 pag:s.
1.
Sumn.ar2:e the proposed SEP 2 tem.
2.
Summarize your conclusions regarding eacn of tne items in.
3.
Provide your recommendations for fur ther f4RC act ion, or indicate if you believe no furthrr action it warranted (other as elected by Regann than routine inspection of the modification, I).
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4.
Provide any additional comments you feel are appropriate.
ag
%\\a es 5
., j',..,It UNITED STATES a
S NUCLEAR REGULATORY COMMISSION L
OFFICE OF INSPECTION AND ENFORCEMENT
%g
,/
Washington, D.C. 20555 INSPECTION AND ENFORCEMENT MANUAL DQASIP 10 CFR 50.59 PART 9800 CFR DISCUSSIONS CHANGES TO FACILITIES, PROCEDURES AND TESTS (OR EXPERIMENTS)
A.
PURPOSE The purpose of this guidance is to clarify the specific 10 CFR 50.59 language relating to the type of proposed changes, tests, or experi-ments that require a record of the safety evaluation specified in 10 CFR 50.59(b).
It is not intended that this guidance delineate specific licensee review criteria which may be used to identify pro-posed changes, tests, or experiments that require a safet evaluation as specified by 10 CFR 50.59(b).
i l
B.
POLICY This' revision to this CFR Discussion does not represent a change in IE policy.
The discussion section has been revised to clarify the application of 10 CFR 50.59 to controls for using, jumpers /lif ted leads and to procedure changes.
Also, the 10 CFR 50.59 flowchart (Appendix
- 1) was updated.
C.
APPLICABILITY:
2515 D.
DISCUSSION 1.
10 CFR 50.59 is composed of three essential parts:
a.
Paragraph (a)(1) is permissive in that it allows the licen-see to make changes to the facility and its operation as described in the Safety Analysis Report (SAR) without prior approval, provided a change in Technical Specifications (TS) is not involved or an "unreviewed safety question" does not exist.
Criteria for determining whether an unreviewed safety question exists are defined in Paragraph (a)(2).
b.
Paragraph (b) requires that the licensee maintain records of changes made under the authority of Paragraph (a)(1).
These records must include a written safety evaluation which pro-vides the basis for determining whether an unreviewed safety question exists.
Paragraph (b) also requires that a report (at least annually) of such changes be submitted to the NRC.
c.
Paragraph (c) requires that proposed changes in Technical Issue Date:
01/01/84 r
CHANGES TO' FAC..oT3ES, PROCEDURES
~
10 CFR 50.59 AND' TESTS (OR EXPERIMENTS)
Specifications be submitted ' to the NRC as an application for
[
license amendment.-
Likewise, proposed changes to the facil-(
ity -or procedures and the proposed conduct of. tests which involve an unreviewed safety question must be submitted to the NRC as an application for license amendment.
2.
It should be noted that the safety evaluation required by 10 CFR 50.59 is only one of the several evaluations and reviews required by the - NRC.
Most. Technical Specifications require that onsite review groups review proposed procedures and modifications or changes to plant equipment or components affecting safety. These review requirements are applicable' whether or not the equipment or component is de. scribed in the SAR.
As a result of the TS required reviews, the need for a safety evaluation to meet 10 CFR 50.59 requirements may be identified.
Appendix' I delineates a l
typical overall review scheme at a facility.
3.
This guidance is to be applied during inspection of facilities holding operating licenses under 10 CFR 50 and is primarily
. directed toward:
Changes.made to those systems and procedures described in a.
2 the SAR, and
~
b.
Performance of tests not described in the SAR.
i
-4.
Within the context of this guidance, any proposed change to a
(
system or procedure as described in the SAR either by text or drawings should be reviewed by the licensee to determine whether it involves an unreviewed safety question.
Changes may involve an unreviewed safety question even though they are "beyond the second isolation valves," or they do not serve a normal safety-
~
related function, since alteration may introduce an unreviewed safety question.
5.
Maintenance activities which do not result in a change to a 4
system (permanent or temporary), or which replace components with replacement parts procured to the same (or equivalent) purchase specification, do not require a written safety evaluation to meet 10 CFR 50.59 requirements.
However, if components described _in the SAR are removed, or their function is altered, or if substi-tute components are utilized, or if changes remain following com-pletion of a maintenance activity, a safety evaluation is re-quired to meet the provisions of 10 CFR 50.59 and the change must be reported to tLe NRC as required by 10 CFR 50.59(b).
6.
In all cases requiring a written safety evaluation, the safety evaluation must provide the basis for determination that the pro-posed change does or does not involve an unreviewed safety ques-tion.
A simple statement of conclusion in itself is not suffi-cient; however, depending upon the significance of the change, the safety evaluation may be quite brief.
Issue Date:
01/01/84.
W
' CfiANGES TO FACILITIE.,, PROCEDURES AND TESTS (OR EXPERIMENTS) 10 CFR 50.59 7.
Listed below are examples of various changes e.o facilities,
(
systems, procedures, and tests which are typical of those requir-ing a 10 CFR 50.59 safety evaluation and those which do not re-quire a safety evaluation under the requirements of 10 CFR 50.59.
a.
Changes in the Facility As Described in the Safety Analysis Report.
This pertains to any changes in the facility which alter the design, function, or method of performing the function of a component, system, or structure described in the SAR.
This would apply to components, systems, and structures described either in the written portion of the SAR or in the drawings contained therein.
Contrasting examples of each case are:
(1) Components.
Replacement of thermocouple in the diesel high-bearing temperature automatic shutdown circuitry (if such a component were described in the SAR) with one made by the same manufacturer, but encompassing different response characteristics, would. require a safety evaluation to meet the requirements of 10 CFR 50.59.
~
On the other hand, replacement of a thermocouple in the diesel high-bearing temperature automatic shutdown circuitry (if such a component were described in the SAR) with one encompassing equivalent response charac-
[
teristics, but made by a different manufacturer, would not require a safety evaluation under the requirements of 10 CFR 50.59.
(2) Systems.
Modifications of the diesel shutdown cir-cuitry (described in the SAR) to provide an automatic
~
i diesel shutdown on high-bearing temperature (shutdown feature not described in application) would require a safety evaluation to meet the requirements of 10 CFR 50.59.
On the other hand, if the methods of initiating automatic diesel shutdown are not described in the SAR, specific automatic shutdown features may be rendered inoperable without the conduct of a safety evaluation under the requirements of 10 CFR 50.59.
(3) Structures.
The erection of a concrete block shield wall within the containment building (shield wall is not described in the SAR) would require a safety evalu-ation to m W the requirements of 10 CFR 50.59.
On the other hand, deletion of a shield wall within the con-tainment building (shield wall not described in the SAR) would not require a safety evaluation under the requirements of 10 CFR 50.59.
(4) Jumpers / Lifted Leads.
Licensee controls over jump-R l;
ers lif ted/ leads should include a docJmented review R process consistent with the one presented in Appendix R Issue Date:
01/01/84
CHANGES TO FACILITIES, PROCEDURES 10 CFR 50.59 (AND TESTS (OR EXPERIMENTS)
If it 's determined that use of a jumper /lif ted R i
?
1.
lead results in a change to the facility as de-R cribed in the SAR and that the resultant change R i
will impact on safety of operation, then a R safety evaluation is required.
This approach R should apply to all types of temporary modifi-R l
cations.
Generally, if a plant system is R l
~
changed by use of jumpers / lifted leads so that R j
it will function differently than described in R the SAR, a safety evaluation would be required. R
{
On the other hand, use of jumpers / lifted leads R that result in plant conditions already analyzed R and approved by NRC would not require a safety R evaluation.
For example, bypassing protection R channels in a manner already described in the R SAR would not constitute an unreviewed safety R question and would not require a safety evalu-R ation under the requirements of 10 CFR 50.59.
R It is expected that only a small percentage of a R j
licensee's jumpers / lifted leads will require a R written safety evaluation R
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b.
Changes in Procedures As Described in the SAR.
This pertains not only to procedures discussed in the initial operations and organizational chapters of the SAR, but also to other pro-cedural-type conaiteents, such as the emergency plan and modes and sequences of plant operation described in the SAR.
If a R procedure results in a deviation from the steps listed in the R SAR or will result in a system operation which deviates from R the way that system is described in the SAR, then a a safety R evaluation should be performed.
Contrasting examples of the R i
above follow.
R (1) If in the description of the radioactive waste system in the SAR, the licensee states that the Shift Supervisor will authorize all radioactive liquid releases, a safety evalu-ation to meet the requirements of 10 CFR 50.59 would be required before assigning this function to another indivi-dual.
On the other hand, if the SAR merely states that radioactive liquid releases will be authorized as detailed by plant procedures, the licensee's predesignation of the authorization function would not require a safety evaluation under the requirements of 10 CFR 50.59.
(2) If the reactor startup procedure, as described in the SAR, contains eight fundamental sequences, the licensee's deci-sion to eliminate one of the sequences would require a safety evaluation to meet the 10 CFR 50.59 requirements.
On the other hand, if the licensee consolidated the eight fundamental sequences but did not alter the basic functions performed, it would not be necessary to conduct a safety evaluation under the requirements of 10 CFR 50.59.
Issue Date:
01/01/84,
o
.d CHANGES'TO FACILITIES, PROCEDURES 10 CFR 50.59 AND TESTS (OR EXPERIMENTS) c.
Conduct Tests and Experiments Not Described in the SAR.
This
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pertains to the performance of an operation not described in the SAR which could have an adverse effect on safety-related systems.
Contrasting examples of such tests or experiments are:
(1) Some plants in the startup testing program have performed a deboration to critical with all rods inserted.
Since this test is performed without deference to the "one stuck rod criterion," a safety evaluation to meet the requirements of 10 CFR 50.59 would be required if the test is not delineated in the SAR.
Since this test may decrease the margin of safety defined in the TS basis, it should, in most in-stances, be classified as an unreviewed safety question.
On the other hand, a test to demonstrate the calibration of the nuclear instrumentation system by perfomance of a secondary plant heat balance would not require a safety evaluation under the requirements of 10 CFR 50.59, even if such a test was not delineated in the SAR, since the test does not in-volve an abnormal mode of operation.
(2) A test to determine if the boric acid evaporator may also be used -for concentration of the steam generator blowdown ef-fluent (function not described in the SAR) would require a safety evaluation to meet the requirements of 10 CFR 50.59, since secondary system chemicals could possibly have a deleterious effect on some components within the reactor
(
coolant pressure boundary. On the other hand, an experiment to determine the decontamination factor of the liquid waste concentrator with. influent activities of 10 2 Ci/ml and 10 5 Ci/ml would not require a safety evaluation under the requirements of 10 CFR 50.59 since such an experiment would not represent departure from normal operational modes.
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d.
General Guidance.
It should be noted that the SARs for a number of older facilities contain floor plans' of onsite buildings that may include trivial detail such as the locating of dividing walls between various offices.
From 'a rigid reading 10 CFR 50.59, it is possible to infer that the removal of a dividing wall between J
two offices constitutes a change from the facility described in 3
the SAR, and therefore requires a: safety evaluation.
- However, 1
the intent of 10 CFR 50.59 is to limit the requirement for writ-ten safety evaluations to facility changes, tests, and experi-ments which could impact the safety of operations.
END l Issue Date:
01/01/84 l
8
__-_-____--L
t a
1
~
CHANGES TO FACILITIES, r ROCEDURES l
10 CFR 50.59 AND TESTS (OR EXPERIMENTS) lCAanoe Proposa1]
Most Tdchnical Specifications (T5) reestre the Onstte Review Group to (1) review all procedures and changes therets that affect nuclear safety, all proposed tests and experiernts that affect nuclear safety.
and all proposed changes to the f acility that affect nuclear safety:
~ ~ " " ~ ~ " " ~ ~ ~ ~ "
3 and (2) to recommend in writing to the Plant Superintendent approval j
or disapproval of these proposals, r
9 Is the Safety Analysis Report (SAR) affected?
(1) Does the proposal change the facility or procedures from their description in the SAR?
{
(2) Does the proposal involve a test or experiment not described in the SAR?
(3) Could the proposal affect nuclear safety in a way not previously evaluated in the SAR?
All answers No Any answer Yes ir lls a enance sn the TS involved?j 10 CTR 50.59 no longer applies..It is still necessary, however, to ask:
go yes Is a change in the T5 involved?
Tes No o
P 4-
~1s an ypreviewec safety question involved?
(1) Is the probability of an occurrence or the conseovences of an accident or malfunction of equipment important to safety previously evaluated in the SAR increased?
(2) is the possibility for an accident or r.alfunction of a different type than any previously evaluated in the SAR created?
(3) 1s the margin of safety as defined in the basis for any technical specification reduced?
Most 15 reovire the Onsite teview Group to render determinations in writing with regard to whether or not the propcsed change 3
constitutes an unreviewed safety ovestion.
(
All answers No <,
Any answer Yes Most T5 require the Of f site Review Group to review proposed changes to procedures, ecutpment or systems, and tests or experiments that involve an unreviewed safety cuestion.
Coc went the change. Include in these
" O P records a written safety evaluation Submit the proposal to tne providing the bases for the determination W for authorization.
that the change, test or expert:nent coes not involve an unreviewed safety ovestion.
l Authorization received.j 1'
l P-ecete with the etaneen a
i l
Most T5 reoutre the Of f site Review Group to review the saf ety evaluations for chan9es to procecures, equipment or systees. and tests or experir.ents s.,_,,,,,,, completed under the provisions of St.59 to verify that such actions did not constitute an vecevsewee safet, ovestion.
Al-1 Issue Date:
01/01/84
_